September 29, 2006 

Mr. David Staudt 

Centers for Disease Control and Prevention 

Acquisition and Assistance Field Branch 

Post Office Box 18070 

626 Cochrans Mill Road – B-140 

Pittsburgh, PA 15236-0295 

Subject: Contract No. 200-2004-03805, Task Order 1: Oak Ridge National Laboratory 

Site Profile Review, SCA-TR-TASK1-0013 

Dear Mr. Staudt: 

SC&A is pleased to submit to NIOSH and the Advisory Board our draft Oak Ridge National 

Laboratory Site Profile Review. Please note that Attachment 2 of this report, Site Expert 

Interview Summary, is currently being reviewed by the Site Experts who participated in the 

interview, and will be forwarded under separate cover as soon as it has been approved. 

If you have any questions or comments on this report, please contact John Mauro at 732-530-

0104. We look forward to discussing this draft report with NIOSH and the Advisory Board. 

Project Manager 

cc: P. Ziemer, PhD, Board Chairperson A. Makhijani, PhD, SC&A 

Advisory Board Members H. Behling, SC&A 

L. Wade, PhD, NIOSH M. Thorne, SC&A 

L. Elliott, NIOSH H. Chmelynski, SC&A 

J. Neton, PhD, NIOSH J. Fitzgerald, Saliant 

S. Hinnefeld, NIOSH J. Lipsztein, SC&A 

L. Homoki-Titus, NIOSH K. Robertson-DeMers, CHP, Saliant 

A. Brand, NIOSH S. Ostrow, PhD, SC&A 

J. Broehm, NIOSH K. Behling, SC&A 

L. Shields, NIOSH T. Bell, Saliant 

Project File (ANIOS/001/13) 

Sincerely, 

John Mauro, PhD, CHP 

1608 SPRING HILL ROAD, SUITE 400 • VIENNA, VIRGINIA • 22182 • 703.893.6600 • FAX 703.821.8236 
Draft 

ADVISORY BOARD ON 

RADIATION AND WORKER HEALTH 

National Institute of Occupational Safety and Health 

Oak Ridge National Laboratory Site Profile Review 

Contract No. 200-2004-03805 

Task Order No. 1 

SCA-TR-TASK1-0013
 

Prepared by 

S. Cohen & Associates 

1608 Spring Hill Road, Suite 400 

Vienna, Virginia 22182 

Saliant, Inc.
 

5579 Catholic Church Road 

Jefferson, Maryland 21755
 

September 2006 

Disclaimer 

This document is made available in accordance with the unanimous desire of the Advisory Board 

on Radiation and Worker Health (ABRWH) to maintain all possible openness in its 

deliberations. However, the ABRWH and its contractor, SC&A, caution the reader that at the 

time of its release, this report is pre-decisional and has not been reviewed by the Board for 

factual accuracy or applicability within the requirements of 42 CFR 82. This implies that once 

reviewed by the ABRWH, the Board’s position may differ from the report’s conclusions. Thus, 

the reader should be cautioned that this report is for information only and that premature 

interpretations regarding its conclusions are unwarranted. 
 

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S. Cohen & Associates: Document No. 

SCA-TR-TASK1-0013 

Technical Support for the Advisory Board on 

Radiation & Worker Health Review of 

Effective Date: 

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NIOSH Dose Reconstruction Program Revision No. 0 – DRAFT 

OAK RIDGE NATIONAL LABORATORY 

SITE PROFILE REVIVEW 

Page 1 of 115 

Task Manager: 

________________________ Date: ___________ 

Joseph Fitzgerald 

Supersedes: 

N/A 

Project Manager: 

________________________ Date: ___________ 

John Mauro, PhD, CHP 
 

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TABLE OF CONTENTS 

Acronyms and Abbreviations ..........................................................................................................5 

1.0 Executive Summary .............................................................................................................9 

1.1 Summary of Stengths .............................................................................................12 

1.2 Summary of Findings.............................................................................................12 

2.0 Scope and Introduction ......................................................................................................16 

2.1 Review Scope.........................................................................................................16 

2.2 Review Approach...................................................................................................18 

2.3 Report Organization...............................................................................................18 

3.0 Assessment Criteria and Methods......................................................................................20 

3.1 Objectives ..............................................................................................................20 

3.1.1	 Objective 1: Completeness of Data Sources .............................................20 

3.1.2	 Objective 2: Technical Accuracy..............................................................20 

3.1.3	 Objective 3: Adequacy of Data.................................................................21 

3.1.4	 Objective 4: Consistency among Site Profiles..........................................21 

3.1.5	 Objective 5: Regulatory Compliance ........................................................21 

4.0	 Site Profile Strengths .........................................................................................................25 

4.1 Completeness of Data ............................................................................................25 

4.2 Adequacy of Data ..................................................................................................25 

4.3 Technical Accuracy/Claimant Favorability ...........................................................26 

4.4 Consistency Among Site Profiles ..........................................................................27 

4.5 Regulatory Compliance .........................................................................................27 

5.0 Vertical Issues ....................................................................................................................28 

5.1 Issue 1: Incomplete Dose Data for the Earlier Years............................................28 

5.1.1	 Internal .......................................................................................................28 

5.1.2	 External ......................................................................................................30 

5.1.3	 Air Sampling ..............................................................................................32 

5.2	 Issue 2: Inadequate Consideration of Missed Dose from Other 

Radionuclides.........................................................................................................32 

5.3	 Issue 3: Problems with Neutron Doses .................................................................37 

5.3.1	 NTA Film Neutron Energy Threshold .......................................................37 

5.3.2	 Neutron-to-Photon Ratios ..........................................................................38 

5.3.3	 Neutron Quality Factors.............................................................................38 

5.3.4	 NTA Film Fading.......................................................................................39 

5.4	 Information Needed Concerning Early Day Dose Data and Selection of 

Workers for Badging..............................................................................................39 

5.4.1	 Dose Data for the Earlier Years .................................................................39 

5.4.2	 Selection of Workers for Badging .............................................................40 
 

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5.5	 Issue 5: Lack of Dose Assignment Procedure for Unmonitored Workers............40 

5.6	 Issue 6: Lack of Data Validation and Verification ...............................................41 

5.6.1	 Data Completeness.....................................................................................41 

5.6.2	 Data Accuracy............................................................................................43 

5.7	 Issue 7: The TBD Fails to Adequately Define and Assess Occupational 

Medical Exposure ..................................................................................................44 

5.8	 Issue 8: Techniques and Protocols Increase Uncertainty of Dose 

Conversion Factors Listed in the TBD ..................................................................46 

5.9	 Issue 9: Frequency and Type of X-ray Exposure is Uncertain. ............................46 

5.10	 Issue 10: Inadequate Consideration of Environmental Dose from
 

Radionuclides Other than I-131 and Tritium. ........................................................47 

5.11	 Secondary Issues....................................................................................................48 

5.11.1 Secondary Issue 1: Potential for High-Fired Oxides ................................48 

5.11.2 Secondary Issue 2: Ingestion Pathway is Not Addressed .........................49 

5.11.3 Secondary Issue 3: Lack of Explanation for Determining MDAs ............49 

5.11.4 Secondary Issue 4: Other Potential Medical Exposures Have Not 

Been Identified...........................................................................................50 

5.11.5 Secondary Issue 5: Additional Factors Contribute to Uncertainties .........50 

5.11.6 Secondary Issue 6: Average Annual Environmental Exposures...............51 

5.11.7 Secondary Issue 7: Inadvertent Ingestion and Inhalation of 

Resuspended Particles Not Considered .....................................................53 

6.0	 Overall Adequacy of the Site Profile as a Basis for Dose Reconstruction ........................54 

6.1	 Satisfying the Five Objectives ...............................................................................54 

6.1.1	 Objective 1: Completeness of Data Sources .............................................54 

6.1.2	 Objective 2: Technical Accuracy..............................................................55 

6.1.3	 Objective 3: Adequacy of Data.................................................................57 

6.1.4	 Objective 4: Consistency Among Site Profiles.........................................57 

6.1.5	 Objective 5: Regulatory Compliance ........................................................59 

6.2	 Usability of Site Profile for Intended Purposes .....................................................59 

6.2.1	 Ambiguous Dose Reconstruction Direction ..............................................60 

6.2.2	 Inconsistencies and Editorial Errors in the Site Profiles............................60 

6.3 Unresolved Policy or Generic Technical Issues ....................................................61 

References......................................................................................................................................63 

Attachment 1: NIOSH Technical Documents Considered During the Review Process...............69
 

Attachment 2: Site Expert Interview Summary............................................................................71 

Attachment 3: Key Questions and Responses for NIOSH/ORAU Regarding Site Profile 

Documents .........................................................................................................................72 

Attachment 4: Summary of Conference Calls on SC&A Questions Provided to NIOSH............90
 

Attachment 5: Consistency Among Site Profiles........................................................................101 
 

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ACRONYMS AND ABBREVIATIONS 

Advisory Board Advisory Board on Radiation and Worker Health 

Al Aluminum 

ANP Aircraft Nuclear Propulsion 

AP Anterior-Posterior 

AEC Atomic Energy Commission 

AMAD Activity Median Aerodynamic Diameter 

Bq Becquerel 

BSR Bulk Shielding Reactor 

CATI Computer-Assisted Telephone Interview 

CEF Critical Experiments Facility 

CEDR Comprehensive Epidemiologic Data Resource 

CEDS Centralized External Dosimetry System 

CER Center for Epidemiologic Research 

CFR Code of Federal Regulations 

Ci Curie 

CXR Chest X-ray 

DCF Dose Conversion Factor 

D&D Decontamination and Decommissioning 

DOSAR Dosimetry Applications Research Calibration Laboratory 

DR Dose Reconstruction 

DOE Department of Energy 

DOELAP Department of Energy Laboratory Accreditation Program 

dpm Disintegrations per Minute 

EC Electronic Capture 

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000 

ESE Entrance Skin Exposure 

FCAF Fuel Cycle Alpha Facility 

GSD Geometric Standard Deviation 

HFIR High Flux Isotope Reactor 

HHIRF Holified Heavy Iron Research Facility 
 

 

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HP Health Physics 

HPRR Health Physics Research Reactor 

HRE Homogeneous Reactor Experiment 

HTO Tritiated Water 

HVL Half Value Layer 

IAAP Iowa Army Ammunition Plant 

ICRP International Commission on Radiological Protection 

IMBA Integrated Modules for Bioassay Analysis 

INEEL Idaho National Engineering and Environmental Laboratory 

IREP Interactive RadioEpidemiological Program 

keV Kilo electron Volt 

LAM Local Air Monitoring 

LANL Los Alamos National Laboratory 

LAT Lateral 

LITR Low Intensity Test Reactor 

LOD Limit of Detection 

μCi Microcurie 

MAC Maximum Allowable Concentration 

MAP Mixed Activation Products 

MCW Mallinckrodt Chemical Works 

MDA Minimum Detectable Activity 

MDC Minimum Detectable Concentration 

MDL Minimum Detectable Dose 

MeV Mega-electron Volt 

MFAP Mixed Fission/Activation Products 

MFP Mixed Fission Products 

mR Milliroentgen 

mrad Millirad 

mrem Millirem 

mrep Millirep 

MSRE Molten Salt Reactor Experiment 

MT Metal Tritides 
 

 

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n-p Neutron-to-photon ratio 

NCRP National Council on Radiation Protection and Measurements 

NIOSH National Institute for Occupational Safety and Health 

NOCTS NIOSH-OCAS Computer Tracking System 

NTA Eastman Kodak Nuclear Track Film Type A 

NTS Nevada Test Site 

NU Natural Uranium 

OBT Organically Bound Tritium 

OCAS Office of Compensation Analysis and Support 

OER Occupational Exposure Record 

ORNL Oak Ridge National Laboratory 

ORR Oak Ridge Research Reactor 

ORAU Oak Ridge Associated Universities 

ORELA Oak Ridge Electron Linear Accelerator 

ORIC Oak Ridge Isochronous Cyclotron 

ORISE Oak Ridge Institute for Science and Education 

OSHA Occupational Safety and Health Administration 

OTIB ORAU Technical Information Bulletin 

OW Open Window 

PA Posterior-Anterior 

PAS Personal Air Sampling 

PCA Pool Critical Assembly 

PFG Photofluorography 

PIC Pocket Ionization Chamber 

POC Probability of Causation 

PROC Procedure 

ppm Parts Per Million 

RaLa Radioactive Lanthanum 

R&D Research and Development 

RCT Radiological Control Technician 

REDC Radiochemical Engineering Development Center 

RFP Rocky Flats Plant 
 

 

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RU Recycled Uranium 

SC&A S. Cohen and Associates 

SRS Savannah River Site 

SRDB Site Research Database 

SSD Source to Surface Distance 

TB Tuberculosis 

TBD Technical Basis Document 

TDF Transuranium Decontamination Facility 

TIB NIOSH Technical Information Bulletin 

TLD Thermoluminescent Dosimeter 

TSR Tower Shielding Reactor 

WISPR NIOSH worker input database 

X-10 Oak Ridge National Laboratory 

Y-12 Plant Y-12 National Security Complex 
 

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1.0 EXECUTIVE SUMMARY 

This report provides the results of an independent audit conducted by S. Cohen and Associates 

(SC&A, Inc.) of the technical basis documents (TBDs) developed by the National Institute for 

Occupation Safety and Health (NIOSH) that make up the site profile for the Oak Ridge National 

Laboratory (ORNL). This audit was conducted during the period February 1, 2006– 

September 15, 2006, in support of the Advisory Board on Radiation and Worker Health 

(Advisory Board) in the latter’s statutory responsibility under the Energy Employees 

Occupational Illness Compensation Program Act (EEOICPA) to conduct such reviews, and 

advise the Secretary of Health and Human Services on the “completeness and adequacy” of the 

EEOICPA program. 

ORNL is located on the Oak Ridge Reservation, along with the Y-12 National Security Complex 

(Y-12) and the Oak Ridge Gaseous Diffusion Plant (K-25). The main laboratory area of ORNL 

is in Bethel Valley. Other ORNL research facilities are located at an adjacent site in Melton 

Valley and at the Y-12 Plant (DOE 1990). The Site Description TBD (Fleming 2006b, pp. 10– 

11) provides the following summary of the scope of the nuclear activities conducted at the 

ORNL: 

Since its operations began in 1943, the mission of Oak Ridge National Laboratory 

(ORNL) has been to conduct research and development (R&D) and production 

missions in support of DOE and its predecessor agencies. Much of the earliest 

site work was devoted to the development and operation of the original plutonium 

production reactor and associated chemical separation facility to test the larger 

production reactors that were being built on the Hanford Site. The Graphite 

Reactor produced gram quantities of plutonium and later fission products [e.g., 

radioactive lanthanum (RaLa)]; other types of radioactive materials were 

separated in other site facilities. Waste control technologies during early site 

operations were in their infancy, and much of the current knowledge of transport 

of radionuclides in the environment was obtained during this time. The ability to 

detect, identify, and quantify radiation types and exposures were progressing 

along with new technologies being discovered in radioisotope production. Much 

of the information gained during the early years at ORNL was used for the design 

of future U.S. Atomic Energy Commission (AEC)/DOE facilities and detection 

systems. Waste radioactive material was released from early site operations as 

gaseous, liquid, and solid effluents with little or no pretreatment. Methods were 

later developed to capture many of the contaminants at their source and to reduce 

overall plant emissions. In some cases, this increased direct exposures to 

individuals in the immediate area and created locations in which incidents and 

spills occurred. 

During the more than 60 years of operations at the site, facilities have been 

constructed, operated, decontaminated, and decommissioned based on need. 
 

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• 	 The operation of the Graphite Reactor for producing plutonium and other 

radioisotopes 

• 	 The development and refinement of chemical processes to separate 

plutonium, uranium, and thorium from irradiated fuel 

• 	 Chemical separation of RaLa from irradiated fuel slugs for use in 

implosion dynamics studies at Los Alamos National Laboratory 

• 	 Operation of facilities for the separation, packaging, and distribution of 

radioisotopes for government and commercial use 

In addition, ORNL developed new reactor technologies. The Laboratory tested 

different reactor designs (pool, pressurized-water, boiling-water, liquid-metal, 

gas-cooled) that were either scrapped or developed further elsewhere. Reactors 

operated at ORNL include the Low-Intensity Test Reactor (LITR), Critical 

Experiments Facility [CEF, at the Y-12 National Nuclear Security Complex 

(Y-12)], Bulk Shielding Reactor (BSR)/Pool Critical Assembly (PCA), Oak Ridge 

Research Reactor (ORR), Tower Shielding Reactor (TSR), Health Physics 

Research Reactor (HPRR), Homogeneous Reactor Experiment (HRE), Aircraft 

Nuclear Propulsion (ANP) Program, the High Flux Isotope Reactor (HFIR), and 

the Molten Salt Reactor Experiment (MSRE). 

SC&A’s review focused on the six TBDs that make up the ORNL site profile. These address 

introduction, site description, occupational internal dose, occupational external dose, 

occupational medical dose, and occupational environmental dose, as they pertain to historic 

radiation exposure of ORNL workers. These TBDs are dated from 2004–2006. As “living” 

documents, TBDs are constantly being revised as new information, experience, or issues arise. 

A complete list of the ORNL TBDs, as well as supporting documents, that were reviewed by 

SC&A is provided in Attachment 1. 

SC&A’s review process included a review of the TBDs, a visit to Oak Ridge, Tennessee, to 

conduct interviews with site experts and identify documents for data retrieval, reviews of 

retrieved ORNL and other historic records, and an exchange of questions and answers, in 

addition to TBD-specific conference calls, between SC&A and its NIOSH and Oak Ridge 

Associated Universities (ORAU) counterparts. The TBDs were evaluated for their 

completeness, technical accuracy, adequacy of data, compliance with stated objectives, and 

consistency with other site profiles, as stipulated in the SC&A Standard Operating Procedure for 

Performing Site Profile Reviews (SC&A 2004). 

The Occupational Medical Dose TBD, ORAUT-TKBS-0012-3 (Fleming 2004), provides little 

documentation to support the assumed techniques, and protocols applied to calculate the dose, 

which is mainly derived from Cardarelli et al. 2002, are accurate. NIOSH believes that when no 

information is readily available about the energy spectrum, it is reasonable to use the 

assumptions for dose conversion factors (DCFs) that are presented in the Implementation Guide. 
 

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The Occupational Environmental Dose TBD (Burns 2004b) appears to rely on emission and 

measurement data; however, it does not indicate the model used for calculations. The TBD 

generally discusses particle size; however, the actual particle size assumptions for assignment of 

internal dose have not been provided. No consideration has been given to the deficiencies in the 

stack and ambient air sampling systems; however, there is a heavy reliance on these systems to 

determine unmonitored worker dose. Exposures considered are limited to I-131, H-3 (starting 

1967), Kr-85, Xe-133, and mixed fission products (MFPs) (optional), while diverse 

radionuclides were handled and potentially released at the site. There has been no consideration 

of potential doses from the release of large uranium particles. Overall, SC&A believes that 

further investigation into environmental source terms and pathways is needed. 

The Occupational External Dose TBD (Burns and Mohrbacher 2004) may result in an 

underestimate of neutron dose. Neutron dose is determined from neutron track emulsion type A 

(NTA) film results and is modified with a correction factor. Some facility-specific neutron 

energy bands are provided; however, in some facilities, the entire spectrum is essentially below 

the practical 1-MeV detection limits of NTA film used in the workers’ badges. From the 

information in current Occupational External Dose TBD, ORAUT-TKBS-0012-6 (Burns and 

Mohrbacher 2004), it is not obvious that the dose reconstructor has sufficient detailed correction 

factors/instructions available to correct for the unmonitored neutron doses resulting from 

neutrons with less than 1 MeV of energy at the numerous facilities at ORNL that produced 

neutron exposures through the years. 

Information available for the dose reconstruction in the early years is limited, inadequate, and in 

some cases, not available. External monitoring for 1943–1944 was limited to the use of pocket 

ionization chambers (PICs), with some experimental badges worn. There is a lack of clear 

discussion on how these monitored and unmonitored doses are derived during this time period. 

Furthermore, the Occupational External Dose TBD (Burns and Mohrbacher 2004) questions the 

validity of these data. The monitoring practices for the years prior to 1951 required further 

investigation to determine if all exposed workers were monitored during this time period. In 

terms of internal exposure, there was an absence of routine internal monitoring until 1949. Early 

bioassay data was limited to plutonium and strontium, although other radionuclides were being 

handled even prior to 1949. 

The Occupational Internal Dose TBD lacks guidance on how to assign dose to radionuclides 

other than transuranics, uranium, activation products, and fission products. As indicated by site 

experts, ORNL handled almost everything on the periodic table at one point or another. There 

has been no screening presented to demonstrate that the secondary radionuclides, particularly 

accelerator- and reactor-produced, are of no dose consequence to the workers. Although ORNL 

handled uranium and radium in the early years, no consideration was given to occupational radon 

exposure. Information was not provided on the activity fractions for plutonium and thorium. 

Activity fractions for plutonium provide critical information for the assessment of dose from 

americium as an impurity. Dose from non-traditional chemical forms of radionuclides, such as 

high-fired oxides and tritides, were not considered. Finally, an adequate rationale for assumption 

of Am-241 in the case of transplutonium bioassay rather than Cm-244 was not provided. 
 

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Issues presented in this report are sorted into the following categories, in accordance with 

SC&A’s review procedures: 

(1) Completeness of data sources 

(2) Technical accuracy 

(3) Adequacy of data 

(4) Consistency among site profiles 

(5) Regulatory compliance 

Following the introduction and a description of the criteria and methods employed to perform the 

review, the report discusses the strengths of the TBD, followed by a description of the major 

issues identified during our review. The issues were carefully reviewed with respect to the five 

review criteria. Several of the issues were designated as primary findings, because they 

represent key deficiencies in the TBDs that need to be corrected, and which have the potential to 

substantially impact at least some dose reconstructions. Others have been designated “secondary 

findings” to both connote their importance for the technical adequacy and completeness of the 

site profile, and to indicate that they have been judged by SC&A to have relatively less influence 

on dose reconstruction or the ultimate significance of worker doses so estimated. 

1.1 SUMMARY OF STENGTHS 

Both the internal and external dose TBDs provided an extensive history of the internal and 

external monitoring program at ORNL. The NIOSH/ORAU team is aware of gaps in the TBDs, 

and has plans to investigate exposure to radon, americium as an impurity, and tritides. Recent 

revisions to the ORNL Site Description TBD captured a number of facilities that were missed in 

the original version of the site description, and provided further information on ORNL processes 

and operations. One benefit of this revision was the inclusion of all buildings mentioned by site 

experts as being missing from the original revision. 

1.2 SUMMARY OF FINDINGS 

Finding 1: Incomplete Dose Data for the Earlier Years. Information available for dose 

reconstruction in the early years is limited, inadequate, or in some cases, not available. External 

beta/gamma monitoring with film badges did not occur until June 1944, while routine neutron 

monitoring was not available until 1949. The neutron dose is reliant on application of a neutronproton (n-p) ratio to the photon dose, yet the TBD questions the dose results from 1943–1944. 

Bioassay was not routinely available prior to 1949, and then only for a few radionuclides. Table 

5A-2, page 39, of the Occupational Internal Dose TBD (Bollenbacher et al. 2006) provides 

minimum detectable activity (MDA) values that have been determined for gross alpha, gross 

beta, and 16 radionuclides found in urinalysis sampling, and gross alpha and 4 radionuclides 

found in fecal sampling. A method for identifying workers and assigning missed dose for those 

potentially exposed to all the assorted radionuclides for which MDAs have been determined 

(Table 5A-2, page 39) is lacking in this document. No consideration was given to early issues 

with significant beta exposures, which caused skin erythema. Consideration of dose from 

uranium particle releases and their subsequent deposition on the skin was not evaluated in the 
 

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TBD. For 1944–1947, the TBD relies on air sampling data; however, very little information is 

provided related to its collection and analysis. Further evaluation should be provided to make 

sure this approach is bounding for unmonitored acute and chronic intakes. 

Finding 2: Inadequate Consideration of Missed Dose from Other Radionuclides. Although 

it acknowledges their existence, the Occupational Internal Dosimetry TBD ORAUT-TKBS0012-5 (Bollenbacher et al. 2006) does not adequately address potential doses from secondary or 

so-called “exotic” radionuclides. The focus of the TBD is on “radionuclides likely to produce a 

measurable internal dose,” including uranium, activation products, fission products, and 

transuranics. Numerous radionuclides were handled at ORNL ranging in quantities from 

fractions of a gram to kilograms. Radionuclides for which co-worker dose is assigned included 

strontium, uranium, plutonium, Am-241, Cs-137, Ce-144, and Ru-106 (Kennedy 2005). 

Potential exposures to reactor- and accelerator-produced radionuclides have not been adequately 

considered. The TBD does not try to ascertain when radionuclides were present, in what 

quantities they were handled, and whether there were suitable methods available for monitoring 

these radionuclides. 

Finding 3: Problems with Neutron Doses. In view of several statements made in the 

Occupational External Dose TBD (Burns and Mohrbacher 2004), the use of NTA film to monitor 

neutron doses at ORNL raises several areas of concern. For example, page 23 of the TBD 

mentions that neutron energy spectra and neutron exposure data before the late 1980s is sparse, 

and that information is particularly lacking for many of the reactors that operated at ORNL early 

in its history. If n-p values are used instead of NTA dose records, these concerns are still valid, 

because using n-p values depend on a detailed knowledge of the gamma and neutron doses, and 

neutron energy spectra at each work location as a function of time. 

Finding 4: Lack of Information Concerning Selection of Workers for Badging. The 

Occupational External Dose TBD (Burns and Mohrbacher 2004) does not provide sufficient 

details concerning who was badged and when, to ensure that workers were sufficiently 

monitored to allow for technically sound dose reconstruction. Page 11 of the TBD states that 

initially only employees required to work in restricted areas more than 3 days per week were 

issued beta-gamma monitoring, and that as late as 1956, there were no strict enforcement policies 

concerning the wearing of monitoring badges. Apparently, the workers that entered restricted 

areas only 1 or 2 days per week did not received badges or any dose of record. Table 6-2, page 

16, of the TBD provides a list of the characteristics of dosimeters from 1944 to present worn by 

radiation workers at ORNL, but does not describe what defined a radiation worker. The TBD 

needs to further refine who was (and was not) monitored and how those selections were made in 

order to be able to determine the adequacy of the dose records. 

Finding 5: Lack of Dose Assignment Procedure for Unmonitored Worker. The 

Occupational External Dose TBD (Burns and Mohrbacher 2004) does not provide a defined 

procedure to assign dose to unmonitored workers. Section 6.5.1 briefly mentions limits of 

detection (LOD) and provides Table 6-24, page 69, listing the LOD and exchange frequency as a 

function of time. However, this should only be applied to the dose missed by the dosimeter worn 

by a worker, not the dose missed because a worker was not badged. This applies to neutron as 
 

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well as photon and beta doses. During these early years, an unmonitored worker could have 

received dose without management or the worker being fully aware of the hazards. The TBD 

needs to provide technically sound dose reconstruction procedures for assigning doses to 

unmonitored workers, especially in the early years (1943–1960s), when radiation hazards were 

not always recognized or effectively addressed. 

Finding 6: Lack of Data Validation and Verification. The validation and verification of the 

data used in dose reconstruction has not adequately been completed. There are indications that 

additional bioassay data exist that are not reflected in the database obtained by ORAU for the 

calculation of MDAs. For example, we became aware that the ORNL has not fully consolidated 

all the occupational exposure records, indicating that some records may not be complete. Also, 

the completeness and accuracy of the external dosimetry data may require further verification to 

ensure field-recorded dose results were integrated into occupational exposure records (OERs). 

This adds to uncertainty of these data. Finally, the environmental air sampling data ratios used in 

the development of co-worker dose from Ru-106, Ce-144, and Cs-137 should be further justified. 

Finding 7: The TBD Fails to Adequately Define and Assess Occupational Medical 

Exposure. The current medical exposure and dose guidelines, as presented in (Kathren 2003), 

go a long way in assuring that all occupational medical exposures are reasonably included in 

determining the overall dose estimations for claimants. Unfortunately, the interpretation to date 

by the contractor (ORAU) has not been applied too conservatively to be claimant favorable. The 

occupational medical dose TBD (Fleming 2006a) assumes an interpretation that also has been 

considered and applied at other sites, such as the Mound Plant and Los Alamos National 

Laboratory (LANL), Paducah, and Pinellas. To this extent, the assumption that medical 

procedures are limited to only one pre-employment chest x-ray and chest x-rays that are part of 

routine physical exams may substantially underestimate worker medical exposure when 

evaluating occupational medical exposure. 

Finding 8: Techniques and Protocols Increase Uncertainty of DCFs listed in the TBD. The 

Occupational Medical Dose TBD (Fleming 2006a) provides little documentation to support the 

assumed techniques and protocols applied to calculate the dose, which is mainly derived from 

NCRP Report 102. The TBD states that a posterior-anterior (PA) chest x-ray was typically the 

only view taken. It is an undocumented assumption in the TBD that exams required only a PA 

view. SC&A has inquired whether definitive protocol existed to validate that chest exams 

possibly included PA views and lateral (LAT) views on a limited basis. NIOSH has 

acknowledged in other TBD reviews that the lack of verifiable protocols is a generic problem at 

many sites, has planned to search all available records, and will include pertinent records and 

references in any future revision of this section of the TBD. The Occupational Medical Dose 

TBD is also deficient in that little documentation exists to validate x-ray protocols, equipment 

maintenance, and upkeep records. 

Finding 9: Frequency and Type of X-ray Exposure is Uncertain. The Occupational Medical 

Dose TBD in Section 3 provides no documentation or references to support the assumption that 

only a limited group of workers received annual x-ray exams after 1970. To the contrary, up 

until about 1985, most DOE sites performed chest x-rays almost on a voluntary basis. DOE 
 

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medical program reviews documented during the early 1990s showed many sites still used chest 

radiography as a general screening exam. Most workers accepted chest x-rays, even though the 

job did not require it. Also, the assumption that workers in special exposure categories, such as 

beryllium workers, were given chest x-rays only as part of their routine physical is not welldocumented and not consistent with special screening guidelines. The TBD applies no 

conservative assumption to cover such exams. 

Finding 10: Inadequate Consideration of Environmental Dose from Radionuclides Other 

than I-131 and Tritium. The Occupational Environmental Dose TBD (Fleming 2006a) focuses 

on onsite airborne I-131 concentration, onsite airborne concentration of MFPs, onsite airborne 

concentrations of tritium, and onsite exposure rate data. Reactors’ releases and waste farms data 

are not adequately considered. 
 

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2.0 SCOPE AND INTRODUCTION 

The review of the Oak Ridge National Laboratory (ORNL) Site Profile was conducted from 

February 1, 2006–September 15, 2006, by a team of SC&A health physicists and technical 

personnel. 

2.1 REVIEW SCOPE 

Under the Energy Employees Occupational Illness Compensation Program Act of 2000 

(EEOICPA) and Federal regulations defined in Title 42, Part 82, Methods for Radiation Dose 

Reconstruction Under the Energy Employees Occupational Illness Compensation Program, of 

the Code of Federal Regulations (42 CFR Part 82), the Advisory Board on Radiation and Worker 

Health (Advisory Board) is mandated to conduct an independent review of the methods and 

procedures used by the National Institute for Occupational Safety and Health (NIOSH) and its 

contractors for dose reconstruction. As a contractor to the Advisory Board, S. Cohen and 

Associates (SC&A, Inc.) has been charged under Task 1 to support the Advisory Board in this 

effort by independently evaluating a select number of site profiles that correspond to specific 

facilities at which energy employees worked and were exposed to ionizing radiation. 

This report provides a review of the following six technical basis documents (TBDs) related to 

historical occupational exposures at ORNL: 

• 	 ORAUT-TKBS-0012-1, Technical Basis Document for Oak Ridge National Laboratory – 

Introduction Rev. 00, August 11, 2004 (Burns 2004a) 

• 	 ORAUT-TKBS-0012-2, Technical Basis Document for Oak Ridge National Laboratory – 

Site Description Rev. 01, August 30, 2006 (Fleming 2006b) 

• 	 ORAUT-TKBS-0012-3, Technical Basis Document for Oak Ridge National Laboratory – 

Occupational Medical Dose Rev. 01 PC-1, July 21, 2006 (Fleming 2006a) 

• 	 ORAUT-TKBS-0012-4, Technical Basis Document for Oak Ridge National Laboratory – 

Occupational Environmental Dose Rev. 00, May 7, 2004 (Burns 2004b) 

• 	 ORAUT-TKBS-0012-5, Technical Basis Document for Oak Ridge National Laboratory – 

Occupational Internal Dose Rev. 00 PC-1, May 30, 2006 (Bollenbacher et al. 2006) 

• 	 ORAUT-TKBS-0012-6, Technical Basis Document for Oak Ridge National Laboratory – 

Occupational External Dose Rev. 00, August 11, 2004 (Burns and Mohrbacher 2004) 

During the course of SC&A’s review, there was a substantial revision of the Site Description 

TBD. This revised TBD provided additional information on the site, including some facilities 

identified by workers as missing from Rev. 0. A minor revision of the Occupational Medical 

Dose TBD was issued at the mid-point of the review. These documents are supplemented by 
 

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technical information bulletins (TIBs), which provide additional guidance to the dose 

reconstructor. A complete list of these documents is available in Attachment 1. 

Implementation guidance is also provided by so-called “workbooks,” which have been 

developed by NIOSH for selected sites to provide more definitive direction to the dose 

reconstructors on how to interpret and apply TBDs, as well as other available information. 

Beyond the conduct of its independent interviews of site experts and former workers, the SC&A 

team is aware of and has requested access to a NIOSH database called “WISPR,” which contains 

NIOSH/ORAU-conducted interviews. It was the team’s understanding that use of the database 

requires training to be provided by ORAU. A formal request has been made for training and 

access to the database. In the meantime, SC&A has reviewed available information from worker 

outreach meetings and public comments during the Advisory Board meetings in Oak Ridge and 

Knoxville, Tennessee. These references are sources of information for the WISPR database. An 

addendum to this report will be provided, based on the results of an evaluation of the information 

found in the WISPR database as necessary. 

SC&A, in support of the Advisory Board, has critically evaluated the ORNL TBDs for the 

following: 

• 	 Determine the completeness of the information gathered by NIOSH in behalf of the site 

profile, with a view to assessing its adequacy and accuracy in supporting individual dose 

reconstructions 

• 	 Assess the technical merit of the data/information 

• 	 Assess NIOSH’s use of the data in dose reconstructions 

SC&A’s review of the six TBDs focuses on the accuracy and completeness of the data that 

characterized the facility and its operations and the use of these data in dose reconstruction. The 

review was conducted in accordance with Standard Operating Procedure for Performing Site 

Profile Reviews (SC&A 2004), which was approved by the Advisory Board. 

The review is directed at “sampling” the site profile analyses and data for validation purposes. 

The review does not provide a rigorous quality control process whereby actual analyses and 

calculations are duplicated or verified. The scope and depth of the review are focused on aspects 

or parameters of the site profile that would be particularly influential in deriving dose 

reconstructions, bridging uncertainties, or correcting technical inaccuracies. This review does 

not explicitly address the issue of radiation exposures to construction, clean-up and 

decommissioning workers, as they are not addressed in the TBDs. 

The six TBDs serve as site-specific guidance documents used in support of dose reconstructions. 

These site profiles provide the health physicists who conduct dose reconstructions on behalf of 

NIOSH with consistent general information and specifications to support their individual dose 

reconstructions. This report was prepared by SC&A to provide the Advisory Board with an 

evaluation of whether and how the TBDs can support dose reconstruction decisions. The criteria 
 

 

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for evaluation include whether the TBDs provide a basis for scientifically supportable dose 

reconstruction in a manner that is adequate, complete, efficient, and claimant favorable. 

Specifically, these criteria were viewed from the lens of whether dose reconstructions based on 

the TBDs would provide for robust compensation decisions. 

The basic principle of dose reconstruction is to characterize the radiation environments to which 

workers were exposed, and determine the level of exposure the worker received in that 

environment through time. The hierarchy of data used for developing dose reconstruction 

methodologies is dosimeter readings and bioassay data, co-worker data and workplace 

monitoring data, and process description information or source term data. 

2.2 REVIEW APPROACH 

SC&A’s review of the TBDs and supporting documentation concentrated on determining the 

comprehensiveness of data collected by NIOSH, the adequacy of existing ORNL personnel and 

environmental monitoring data, and the evaluation of key dose reconstruction assumptions. 

2.3 REPORT ORGANIZATION 

In accordance with directions provided by the Advisory Board and with site profile review 

procedures prepared by SC&A and approved by the Advisory Board, this report is organized into 

the following sections: 

(1) Executive Summary 

(2) Scope and Introduction 

(3) Assessment Criteria and Methods 

(4) Site Profile Strengths 

(5) Vertical Issues and Secondary Issues 

(6) Overall Adequacy of the ORNL Site Profile as a Basis for Dose Reconstruction 

Based on the issues raised in each of these sections, SC&A prepared a list of findings, which are 

provided in the Executive Summary. Issues are designated as findings if SC&A believes that 

they represent deficiencies in the TBD that need to be corrected and which have the potential to 

have a substantial impact on at least some dose reconstructions. Issues can also be designated as 

Secondary Issues if they simply raise questions, which, if addressed, would further improve the 

TBDs and may possibly reveal deficiencies that will need to be addressed in future revisions of 

the TBDs. 

Many of the issues that surfaced in the report correspond to more than one of the major 

objectives (i.e., strengths, completeness of data, technical accuracy, consistency among site 

profiles, and regulatory compliance). Section 6.0 provides in summary form a list of the issues, 

and to which objective the particular issue applies. Attachment 5 provides a more in-depth 

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In many ways, the TBDs have done a successful job in addressing a series of technical 

challenges. In other areas, the TBDs exhibit shortcomings that may influence some dose 

reconstructions in a substantial manner. Major issue areas include the following: 

• 	 Insufficient data for early worker dose reconstructions 

• 	 Inadequate consideration of missed dose from secondary radionuclides and radionuclide 

impurities 

• 	 Occupational exposure to radon is not fully addressed in the TBDs 

• 	 Insufficient characterization of the monitored workforce for beta/gamma and neutron 

monitoring 

• 	 Underestimation of neutron dose when using NTA file and in developing n-p ratios 

where the initial photon dose is not well characterized 

• 	 Lack of external dose assignment methodology for unmonitored workers 

• 	 Incomplete evaluation of medical x-ray exposures, especially in the early years 

• 	 Inadequate validation of source data used in dose reconstruction 

• 	 Concerns with reported deficiencies of air monitoring sampler locations at the site 

SC&A believes that these important issues need to be effectively dealt with in any upcoming 

revisions to the ORNL site profile TBDs in order that more claimant-favorable dose 

reconstructions can be effectively conducted in areas where these data gaps exist. 
 

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3.0 ASSESSMENT CRITERIA AND METHODS 

SC&A is charged with evaluating the approach set forth in the site profiles that is used in the 

individual dose reconstruction process. These documents are reviewed and evaluated for their 

completeness, technical accuracy, adequacy of data, consistency with other site profiles, and 

compliance with the stated objectives, as defined in SC&A Standard Operating Procedure for 

Performing Site Profile Reviews (SC&A 2004). This review is specific to the ORNL Site Profile 

and supporting TIBs; however, items identified in this report may be applied to other facilities, 

especially facilities with similar source terms and exposure conditions. The review identifies a 

number of issues and discusses the degree to which the site profile fulfills the review objectives 

delineated in SC&A’s site profile review procedure. 

3.1 OBJECTIVES 

SC&A reviewed and evaluated the site profile with respect to the degree to which technically 

sound judgments or assumptions are employed. In addition, the review identifies assumptions by 

NIOSH that give the benefit of the doubt to the claimant. 

3.1.1 Objective 1: Completeness of Data Sources 

This objective requires SC&A to identify principal sources of data and information that are 

applicable to the development of the site profile. The two elements examined under this 

objective include (1) determining if the site profile made use of available data considered 

relevant and significant to the dose reconstruction, and (2) investigating whether other 

relevant/significant sources are available, but were not used in the development of the site 

profile. For example, if data are available in site technical reports or other available site 

documents for particular processes, and if the TBDs have not taken into consideration these data 

where it should have, this would constitute a completeness of data issue. The ORAU site profile 

document database, including the referenced sources in the TBDs, was evaluated to determine 

the relevance of the data collected by NIOSH to the development of the site profile. 

Additionally, SC&A evaluated records captured at ORNL as a part of their records retrieval 

effort, and publicly available records. 

3.1.2 Objective 2: Technical Accuracy 

This objective requires SC&A to perform a critical assessment of the methods used in the site 

profile to develop technically defensible guidance or instructions, including evaluating field 

characterization data, source term data, technical reports, standards and guidance documents, and 

literature related to processes that occurred at ORNL. The goal of this objective is to first 

analyze the data according to sound scientific principles, and then to evaluate this information in 

the context of compensation. If, for example, SC&A found that the technical approach used by 

NIOSH was not scientifically sound or claimant favorable, this would constitute a technical 

accuracy issue. 
 

 

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3.1.3 Objective 3: Adequacy of Data 

Objective 3 requires SC&A to determine whether the data and guidance presented in the site 

profile are sufficiently detailed and complete to conduct dose reconstruction, and whether a 

defensible approach has been developed in the absence of data. In addition, this objective 

requires SC&A to assess the credibility of the data used for dose reconstruction. The adequacy 

of the data identifies gaps in the facility data that may influence the outcome of the dose 

reconstruction process. For example, if a site did not monitor all workers exposed to neutrons 

who should have been monitored, this would be considered a gap and, thus, an inadequacy in the 

data. 

3.1.4 Objective 4: Consistency among Site Profiles 

This objective requires SC&A to identify common elements within site profiles completed or 

reviewed to date, as appropriate. In order to accomplish this objective, the ORNL TBD was 

compared to several of the sites already reviewed by SC&A. A detailed analysis of this review is 

provided in Attachment 5. 

3.1.5 Objective 5: Regulatory Compliance 

The Regulatory Compliance requires SC&A to evaluate the degree to which the site profile 

complies with stated policy and directives contained in 42 CFR Part 82. In addition, SC&A 

evaluated the TBD for adherence to general quality assurance policies and procedures utilized 

for the performance of dose reconstructions. In order to place the above objectives into the 

proper context as they pertain to the site profile, it is important to briefly review key elements of 

the dose reconstruction process, as specified in 42 CFR Part 82. Federal regulations specify that 

a dose reconstruction can be broadly placed into one of three discrete categories. These three 

categories differ greatly in terms of their dependence on and the completeness of available dose 

data, as well as on the accuracy/uncertainty of data. 

Category 1: Least challenged by any deficiencies in available dose/monitoring data are dose 

reconstructions for which even a partial assessment (or minimized dose(s)) corresponds to a 

probability of causation (POC) value in excess of 50%, and assures compensability to the 

claimant. Such partial/incomplete dose reconstructions with a POC greater than 50% may, in 

some cases, involve only a limited amount of external or internal data. In extreme cases, even a 

total absence of a positive measurement may suffice for an assigned organ dose that results in a 

POC greater than 50%. For this reason, dose reconstructions under this category may only be 

marginally affected by incomplete/missing data or uncertainty of the measurements. In fact, 

regulatory guidelines recommend the use of a partial/incomplete dose reconstruction, the 

minimization of dose, and the exclusion of uncertainty for reasons of process efficiency, as long 

as this limited effort produces a POC of greater than or equal to 50%. 

Category 2: This category adopts the “worst-case” assumptions in dose reconstruction in order 

to derive maximal or highly improbable dose assignments. For example, a worst-case 

assumption may place a worker at a given work location 24 hours per day and 365 days per year. 
 

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The use of such maximized (or upper-bound) values, however, is limited to those instances 

where the resultant maximized doses yield POC values below 50%, which are not compensated. 

For this second category, the dose reconstructor needs only to ensure that all potential internal 

and external exposure pathways have been considered. 

The obvious benefit of “worst-case” assumptions and the use of maximized doses in dose 

reconstruction is efficiency, which is achieved by the fact that maximized doses avoid the need 

for precise data and eliminates consideration for the uncertainty of the dose. Lastly, the use of 

bounding values in dose reconstruction minimizes any controversy regarding the decision not to 

compensate a claim. 

Although simplistic in design, to satisfy this type of a dose reconstruction, the TBD must, at a 

minimum, provide information and data that clearly identify (1) all potential radionuclides, 

(2) all potential modes of exposure, and (3) upper limits for each contaminant and mode of 

exposure. Thus, for external exposures, maximum dose rates must be identified in time and 

space that correspond to a worker’s employment period, work locations, and job assignment; 

similarly, in order to maximize internal exposures, highest air concentrations and surface 

contaminations must be identified. 

Category 3: This represents the most complex and challenging dose reconstruction category. It 

consists of claims where the case cannot be dealt with under one of the two categories above. 

For instance, when a minimum dose estimate does not result in compensation, a next step is 

required to make a more complete estimate. Or when a “worst-case” dose estimate that has 

assumptions that may be physically implausible results in a POC greater than 50%, a more 

refined analysis is required. A more refined estimate may be required either to deny or to 

compensate. In such dose reconstructions, which may be represented as “reasonable,” NIOSH 

has committed to resolve uncertainties in favor of the claimant. According to 42 CFR Part 82, 

NIOSH interprets “reasonable estimates” of radiation dose to mean the following: 

. . . estimates calculated using a substantial basis of fact and the application of 

science-based, logical assumptions to supplement or interpret the factual basis. 

Claimants will in no case be harmed by any level of uncertainty involved in 

their claims, since assumptions applied by NIOSH will consistently give the 

benefit of the doubt to claimants. [Emphasis added.] 

In order to achieve the five objectives described above, SC&A reviewed each of the six TBDs, 

their supplemental attachments, and TIBs, giving due consideration to the three categories of 

dose reconstructions that the site profile is intended to support. The six ORNL TBDs provide 

well-organized information for the dose reconstructor when adequate data were available to do 

that comprehensively. 

ORAUT-TKBS-0012-1, Rev. 00, Technical Basis Document for Oak Ridge National Laboratory 

– Introduction (Burns 2004a), explains the purpose and the scope of the site profile. SC&A was 

attentive to this section, because it explains the role of each TBD in support of the dose 

reconstruction process. During the course of its review, SC&A was cognizant of the fact that the 

site profile is not required by the EEOICPA or by 42 CFR Part 82, which implements the statute. 
 

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Site profiles were developed by NIOSH as a resource to the dose reconstructors for identifying 

site-specific practices, parameter values, and factors that are relevant to dose reconstruction. 

Based on information provided by NIOSH personnel, SC&A understands that site profiles are 

living documents, which are revised, refined, and supplemented with TIBs as required to help 

dose reconstructors. Site profiles are not intended to be prescriptive nor necessarily complete in 

terms of addressing every possible issue that may be relevant to a given dose reconstruction. 

Hence, the introduction helps in framing the scope of the site profile. As will be discussed later 

in this report, NIOSH may want to include additional qualifying information in the introduction 

to this and other site profiles describing the dose reconstruction issues that are not explicitly 

addressed by a given site profile. 

ORAUT-TKBS-0012-2, Rev. 01, Technical Basis Document for Oak Ridge National Laboratory 

– Site Description (Fleming 2006b), is an extremely important document, because it provides a 

description of the facilities, processes, and historical information that serve as the underpinning 

for subsequent ORNL TBDs. 

ORAUT-TKBS-0012-3, Rev. 01 PC-1, Technical Basis Document for Oak Ridge 

National Laboratory – Occupational Medical Dose (Fleming 2006a), provides an 

overview of the sources, types of exposure, and the frequency of exams that workers 

potentially received. 

ORAUT-TKBS-0012-4, Rev. 00, Technical Basis Document for Oak Ridge National Laboratory 

– Occupational Environmental Dose (Burns 2004b), provides background information and 

guidance to dose reconstructors for reconstructing the doses to unmonitored workers outside of 

the facilities at the site who may have been exposed to routine and episodic (accidental) airborne 

emissions from these facilities. 

ORAUT-TKBS-0012-5, Rev. 00 PC-1, Technical Basis Document for Oak Ridge National 

Laboratory – Occupational Internal Dose (Bollenbacher et al. 2006), presents background 

information and guidance to dose reconstructors for deriving occupational internal doses to 

workers. 

ORAUT-TKBS-0012-6, Rev. 00, Technical Basis Document for Oak Ridge National Laboratory 

– Occupational External Dose (Burns and Mohrbacher 2004), presents background information 

and guidance to dose reconstructors for deriving occupational external doses to workers. 

Site expert interviews were conducted from February 22– March 3, 2006, in Oak Ridge, 

Tennessee, with former and current ORNL employees; Bechtel-Jacobs, Inc., employees; and 

Department of Energy–Oak Ridge Operations Office oversight personnel. The purpose of these 

interviews was to receive first-hand accounts of past radiological control and personnel 

monitoring practices at ORNL, and better understand how operations were conducted. 

Interviewees were selected to represent a reasonable cross-section of production areas and job 

categories. References to specific site experts have been omitted for privacy reasons. The 

individuals were given the opportunity to review their interview summary for accuracy. This is 

an important safeguard against missing key issues or misinterpreting some vital piece of 

information. To ensure that classified information had not been included in the interview notes, 
 

 

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the notes were reviewed by a classification officer prior to release to SC&A. The compilation of 

all site expert interviews is provided in Attachment 2. 

In accordance with SC&A’s site profile review procedures, SC&A performed an initial review of 

the six TBDs, their supporting documentation, and the two TIBs directly related to the ORNL 

site. SC&A submitted written questions to NIOSH pertaining to the ORNL Site Profile on 

August 1, 2006. NIOSH/ORAU provided written responses to these questions on August 25, 

2006, prior to a conference call with SC&A. The questions and responses are provided in 

Attachment 3. 

A conference call was held on September 6, 2006, with NIOSH/ORAU to allow them to provide 

clarifications and to explain the approaches employed in the site profile TBDs. A summary of 

this conference call is provided in Attachment 4. 

An extensive comparison was done between the methodologies used in the ORNL TBD 

and other TBDs reviewed to date to determine environmental, internal, and external 

doses. This comparison focused on the methodologies and assumptions associated with 

dose reconstruction and resultant values used to obtain a POC. A detailed analysis is 

provided in Attachment 5. 

Information provided in writing and as a part of the conference call with NIOSH was evaluated 

against the preliminary findings to finalize the vertical issues1

 addressed in the audit report. 

There are two levels of detailed review for this report. First, SC&A topical expert members 

review the report internally. This is then followed by what is referred to as the expanded review 

cycle, which will consist of a review of this draft by the Advisory Board and NIOSH. The first 

one of these has been completed. 

A matrix containing the summary of findings is prepared by SC&A and released either 

simultaneously or shortly after the issuance of the review. After the Advisory Board and NIOSH 

have an opportunity to review this draft, a Working Group is formed by the Advisory Board to 

track issues identified within the review to their conclusion. NIOSH/ORAU is allowed to 

respond to the findings. In turn, SC&A is provided with the responses and evaluates them prior 

to meeting in person with the Working Group. During these Working Group meetings, both 

parties may decide the item has been resolved and is closed, or that additional action and/or 

clarification is needed. This process continues until all issues have been resolved. Reviews are 

also published on the NIOSH Web site and discussed at the next Advisory Board meeting. 

Finally, it is important to note that SC&A’s review of the six TBDs and their supporting TIBs is 

not exhaustive. These are large, complex documents, and SC&A used its judgment in selecting 

those issues that we believe are important with respect to dose reconstruction. 

1

 The term “vertical issues” refers to specific issues identified during our review, which were identified as 

requiring more in-depth analysis due to their potential to have a significant impact on dose reconstruction. 
 

 

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4.0 SITE PROFILE STRENGTHS 

In developing a TBD, the assumptions used must be fair, consistent, and scientifically robust, 

and uncertainties and inadequacies in source data must be explicitly addressed. The 

development of the TBD must also consider efficiency in the process of analyzing individual 

exposure histories, so that claims can be processed in a timely manner. With this perspective in 

mind, we identified a number of strengths in the ORNL site TBDs. These strengths are 

described in the following sections. 

4.1 COMPLETENESS OF DATA 

In an effort to comprehensively address the range of facilities and processes at ORNL, NIOSH 

effectively compiled facility-specific information on major facilities and processes. Descriptions 

were provided for 78 facilities and 6 major processes (e.g., accelerators, reactors). A 

comprehensive and effective overview of key operations, their location, and dates of operation is 

also provided in the site profile. SC&A considers this an important start in providing 

background information for dose reconstructors. In fact, the ORNL Site Description TBD 

underwent a substantial rewrite during the course of the review. In site expert interviews 

conducted by SC&A, workers mentioned several facilities that were not included in Revision 0 

of the site description. Revision 1 of the site description captured these facilities. This not only 

provides a more complete document, but shows workers they are being heard. 

In developing the site profile, NIOSH/ORAU drew upon information contained in 226 reports 

cited in the reference sections. Process information was drawn from the Oak Ridge Dose 

Reconstruction project, historical program overviews, technical documents, and individuals 

currently or formerly involved in the Radiological Control group at ORNL. NIOSH/ORAU met 

with the ORNL Labor and Trades Council on November 8, 2004, in order to identify worker 

concerns and discuss the TBDs. This interaction has helped to provide valuable insight into site 

operations and processes. In addition, the issuances of several TIBs reflect ongoing efforts by 

NIOSH to continually improve guidance provided to dose reconstructors. 

4.2 ADEQUACY OF DATA 

The TBDs benefited from having access to information and data that were compiled as a part of 

the ORNL programs, as follows: 

(1) Radiological control personnel have implemented improved procedures and technologies 

over time to reduce radiation dose to workers, and have improved personnel monitoring 

programs. 

(2) Historical documentation of the Radiological Control program, including research 

activities, is well documented. 

(3) ORNL implemented environmental monitoring, including stack monitoring, in-perimeter 

monitoring, offsite monitoring, and ambient external exposure monitoring. 
 

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(4) Starting in 1951, dosimeters were coupled with security badges, which helped ensure all 

workers were monitored for external radiation. 

A good breakdown of the various site locations and activities as a function of time was provided. 

Dosimetry methods used for beta, photon, and neutron dose monitoring were described 

separately, and also as a function of time and technology changes. Some prescribed health 

physics monitoring procedures and record-keeping methods were described, along with the 

logbooks containing this information. ORNL has been fortunate to have noted scientists and 

research and development (R&D) facilities to address radiation and health physics problems 

throughout its years of operations. 

The Occupational Medical Dose TBD (Fleming, 2006a) outlines the operating parameters and 

assumptions used in the development of x-ray dose in Table 3-3 (pg. 18). This summary is very 

helpful in succinctly presenting information to the dose reconstructor. 

Site experts commented that the Occupational Environmental Dose TBD (Burns 2004b) appears 

to have utilized the correct references for evaluating the historical program. 

4.3 TECHNICAL ACCURACY/CLAIMANT FAVORABILITY 

The Occupational External Dose TBD (Burns and Mohrbacher 2004) was reviewed, and several 

areas of strengths were identified that will be useful in performing/evaluation dose 

reconstructions: 

• 	 Useful information concerning beta-gamma fields was provided (pp. 22–23). 

• 	 Information concerning many workplace neutron and photon radiation fields 

characterized in 1989–1991 was provided (pp. 23–44). 

• 	 Thermoluminescent Dosimeter (TLD) neutron and gamma dose data for 1990–2004 for 

numerous facilities and work groups was provided (pp. 44–60). 

• 	 Attachments 6A–6F provide a good summary of beta-gamma and neutron 

characterization and adjustment factors that is helpful in performing dose reconstructions 

without having to search through the complete TBD. 

The Occupational External Dose TBD (Burns and Mohrbacher 2004) is fairly well written, and 

easy to follow. It incorporates a number of essential elements (informative table of contents, 

chronological flow of information, summary tables, etc.) sometimes only found in later revisions 

of TBDs. 

The Occupational Internal Dose TBD (Bollenbacher et al. 2006) provides a good description of 

the in-vivo and in-vitro monitoring history at ORNL. There is a concerted effort to 

independently evaluate minimum detectable concentrations (MDCs) for bioassay techniques 

based on analytical records. The TBD recognizes the shortcomings in the in-vivo monitoring 
 

 

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program and takes this into consideration. NIOSH/ORAU has concurred with SC&A that 

potential radon, tritide, and americium exposure from plutonium mixtures requires further 

investigation. 

4.4 CONSISTENCY AMONG SITE PROFILES 

Dose assignments to Y-12 Plant workers and ORNL workers located at Y-12 is not equivalent 

for the same or similar operations. Consideration should be given to the radiological conditions 

at this location and how they may impact dose. 

The implementation of ORAUT-OTIB-0006, Dose Reconstruction for Occupationally Related 

Diagnostic X-Ray Procedures, Revision 2 (Kathren 2003), has provided a great deal of 

consistency between different site profile assumptions. Incorporation of Revision 3 PC-1 

(Kathren and Shockley 2005) of this document would provide further consistency among site 

profiles. The NIOSH/ORAU team is moving in the right direction to obtain consistency in most 

assumptions for occupational medical exposure. 

4.5 REGULATORY COMPLIANCE 

The TBDs’ use of personnel monitoring data and environmental monitoring data to determine 

dose is consistent with the requirements outlined in 42 CFR Part 82, as follows: 

• 	 Where in-vivo and in-vitro analyses are available, this information is provided for use in 

determination of internal dose. 

• 	 Where routine beta/gamma and neutron dosimeters are available and adequate, this 

information is provided for use in determination of external exposure. 

• 	 Where environmental measurements are available, these data are used as the basis for 

environmental dose. 
 

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5.0 VERTICAL ISSUES 

SC&A has developed a list of key issues regarding the ORNL Site Profile. These issues relate to 

each of the five objectives defined in SC&A’s review procedures (SC&A 2004). Some issues 

are related to a particular objective, while others cover several objectives. Many of the issues 

raised below are applicable to other DOE and Atomic Weapons Employer sites, and should be 

considered in the preparation and revision of other site profiles. 

5.1 ISSUE 1: INCOMPLETE DOSE DATA FOR THE EARLIER YEARS 

Information available for dose reconstruction in the early years is limited, inadequate, or in some 

cases, not available. External monitoring for the years 1943–1944 was limited to pocket 

ionization chambers (PICs). Neutron dosimetry was not available until 1949 when ORNL 

implemented NTA film. The lack of complete photon dosimetry data in the early days brings 

into question the use of n-p ratios for the assignment of neutron doses. Bioassay processes were 

not routinely available prior to 1949, and the first bioassay techniques were limited to Sr-90 and 

plutonium. Reactor, accelerator, and chemical separations operations began at ORNL prior to 

the implementation of the routine bioassay program. Prior to this time, the reliance was on air 

sampling and contamination control, which forms the basis for dose reconstruction from 1943– 

1947. The production and handling of other radionuclides such as Ba/La-140, radium, and other 

alpha and beta emitters, preceded many of the bioassay techniques for these sources. Overall, 

the Health Physics group of the 1940s focused their efforts on keeping individuals within the 

defined tolerances of the time, rather than eliminating internal exposure all together. 

The Graphite Reactor, built to produce plutonium for the weapons program, went critical on 

November 4, 1943. Process separations at the plutonium-separation pilot plant began in 

December 1943, with uranium irradiated at an accelerator at Washington University. The first 

shipment of separated plutonium was sent from Oak Ridge to Los Alamos in February 1944 

(Johnson and Schaffer 1994). In 1944, Los Alamos requested that ORNL produce 100 Ci of 

Ba-140, which decayed to La-140, for use in weapons development. In 1944, the Graphite 

Reactor was used to irradiate Bismuth-210 slugs for the production of Po-210 used in initiators. 

Other early ORNL missions were to perform R&D related to nuclear fission, nuclear reactors, 

and the plutonium-separation process. The Chemistry Division supported not only the 

plutonium-separation operations, but preparation and use of radionuclide tracers (e.g., tritium, 

radioactive phosphorus, and radioactive antimony). Analytical techniques were developed for 

the identification of fission products. The Health Division conducted animal experiments to 

evaluate radiation effects. Around mid-1945, the laboratory began to separate U-233 from 

irradiated thorium. In August 1946, radionuclides produced at ORNL became available to those 

outside the Manhattan Project (Quist 2000). 

5.1.1 Internal 

Table 5A-2, page 39, of the Occupational Internal Dose TBD (Bollenbacher et al. 2006) provides 

minimum detectable activity (MDA) values that have been determined for gross alpha, gross 

beta, and 16 radionuclides found in urinalysis sampling, and gross alpha and 4 radionuclides 
 

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found in fecal sampling. A method for identifying workers and assigning missed dose is lacking 

in this document for those potentially exposed to all the assorted radionuclides, for which MDAs 

have been determined (Table 5A-2, page 39). Likewise, for all the other radionuclides, such as 

those produced by the ORNL Isotopes Production Group at Y-12, a method is lacking for 

assigning a missed dose. This issue needs to be addressed regarding this group of workers who 

received considerable attention during the Y-12 Special Exposure Cohort (SEC) petition review. 

NIOSH has concluded that the radionuclides likely to produce a measurable internal dose were 

uranium, transuranics, activation products, and fission products. As mentioned in the 

introduction to this section, there was potential exposure to uranium, plutonium, fission products, 

Ba/La-140, Po-210, tritium, U-233, and thorium starting in the 1940s. Table 5-9 provides invitro MDAs for 1947 to 1989. Exposures to even those radionuclides of concern, as defined by 

NIOSH, occurred prior to 1947 (Bollenbacher et al. 2006, pg. 16). Furthermore, isotope-specific 

in-vitro analysis did not become routine until 1989, with the radionuclide being determined in 

some cases based on process knowledge. Page 11 of the ORNL Occupational Internal Dose 

TBD ORAUT-TKBS-0012-5 (Bollenbacher et al. 2006) notes the following: 

...urine samples were collected in the early years of the bioassay program based 

on the area health physicist’s knowledge of field conditions (e.g., known 

spills/incidents, air and contamination sample results, etc.). This practice of 

scheduling did not utilize a specified sampling frequency. 

On page 21, the TBD further notes the following: 

Although several located documents stated that baseline and specified monitoring 

frequencies were utilized to make in-vivo measurements, Berger (2003) and 

McLaughlin (2004) indicated that a full in-vivo monitoring program did not exist 

at ORNL until approximately 1994, when site internal dosimetrists became 

responsible for identifying personnel for counting. 

This practice for determining sampling for both in-vitro and in-vivo monitoring 

throughout the history of the site through the 1980s or even 1990s raises a serious 

question regarding unmonitored workers with potential for uptakes. The issue is further 

exacerbated in Table 5-3 on page 11, where it states that these sampling frequencies 

should be followed; however, the text notes (Section 5.1.4.1) that these may not have 

been followed. (Emphasis added.) 

With the diversity of radionuclides handled by ORNL (including operations at Y-12), it is 

uncertain how NIOSH/ORAU has determined that radionuclides other than transuranics, 

activation products, fission products, and uranium are the only significant dose contributors. It is 

also unclear what the TBD means by “activation product.” Are these exclusively those 

radionuclides created by activation of material in reactors, or does it include accelerator 

material? A screening analysis should be completed for all radionuclides mentioned in the Site 

Description TBD to ensure there was a method for detection of this radionuclide while it was 

handled, or that the dose consequences for the radionuclide are trivial. 
 

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5.1.2 External 

With the operation of a reactor and the handling of irradiated fuel, the potential existed not only 

for beta/gamma exposure, but for neutron exposure as well. The Occupational External Dose 

TBD states the following (Burns and Mohrbacher 2004): 

Beginning in October 1943, ORNL issued paired pocket dosimeters to workers 

assigned to duties in the restricted areas associated with Graphite Reactor 

operations. Pocket dosimeters were used in pairs because of reliability issues 

associated with early designs. ORNL issued two dosimeters so there was a back 

up if one of them failed, went off the scale, or was otherwise unusable. An 

individual received these dual dosimeters if restricted area entries were likely 

more than three times per week. If entry into restricted areas was likely to be less 

than three times per week or on a random basis, workers received dosimeters for 

each entry. 

The TBD further indicates that experimental dosimeters were worn by some personnel for the 

first half of 1944 (Burns and Mohrbacher 2004). The Occupational External TBD does not 

provide instruction on how to assign dose for 1943 and 1944 to workers who may have only 

worn PICs. The error associated with these units will differ from that of a film badge and should 

be considered. Co-worker doses are provided in OTIB-0021 (Kennedy 2005) for this period; 

however, the Occupational External Dose TBD makes the following statement (Burns and 

Mohrbacher 2004): 

Film dosimeter readings (results) encountered in ORNL personnel monitoring 

histories before June 25, 1944, could be unreliable because the use of dosimeters 

during this time was experimental (Hart 1966). 

Information on how this experimental data is being used in both individual and co-worker doses 

is minimal. There is no information on which portion of the population wore these dosimeters, 

and if they were representative of the population. The policy for monitoring from June 1944 

through November 1951 was based on the number of entries an individual made into a radiation 

area, or whether they worked with radioactive material. Routine dosimeters were assigned to 

those who frequented radiation areas, while trip dosimeters were assigned to individuals who 

intermittently visited radiation areas. The uncertainty related to the photon monitoring system 

does not instill confidence in the reconstruction of dose for the early years. 

The Occupational External Dose TBD indicates that neutron monitoring was formally introduced 

in 1949 (Burns and Mohrbacher 2004): 

Neutron personnel monitoring using NTA film was formally introduced at ORNL 

around 1949. However, NTA film was apparently used on a limited basis to 

supplement field measurements as early as February 1945 (Wirth, Morgan, and 

Curtis 1945).... Once it became available, NTA film was used in all assigned 
 

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employee badges and exchanged on a weekly basis. However, the film was not 

processed unless the health physicist recommended it. 

Bradley (1945) indicated that routine monitoring of 105 personnel was conducted by inserting a 

special film in the regular beta/gamma meter. Neutron monitoring was initially determined by 

the Health Physics Division Director, as indicated by Ledbetter (1948) and later by Health 

Physicists in the field. 

NIOSH uses primarily n-p ratios to determine the external dose. Currently, they have developed 

n-p ratios from neutron dosimetry results for 1999–2004. Neutron ratios vary substantially, 

depending on how they are generated and what shielding is present. Radiological conditions 

today are significantly different from those from the 1940s in terms of implementation of 

engineering and administrative controls to limit dose. ORNL currently has a research mission 

whereas in the 1940s, the focus was on production of material for the weapons program. NIOSH 

needs to further explain how the application of n-p ratios from the present time is representative 

of those from the 1940s. Secondly, the uncertainty related to the photon monitoring system does 

not instill confidence in the reconstruction of dose for the early years. 

The Laboratory was involved in activities that led to significant beta exposures. For example, 

Morgan and Peterson (1999) state the following: 

Skin erythema from beta radiation represented a special early concern at X-10 

because some employees refused to wear heavy leaded gloves and even handled 

uranium slugs with their bare hands when we weren’t looking. 

Pocket ionization chambers are designed to measure photon exposure, rather than beta exposure. 

For those years when PICs were the exclusive monitoring system, there was an absence of 

personnel beta monitoring. Potential skin exposure from environmental releases was also a 

concern with early particle emissions that deposited on the ground and, in some cases, on 

personnel. 

The other serious incident in the very early period involved the release of small 

uranium particles that contained plutonium and fission products. These “hot” 

particles from a chemical separation operation would fall all over the laboratory 

areas (Morgan and Peterson 1999). 

In the conference call held with NIOSH, it was indicated that these particles would be measured 

by the ambient air monitoring system (see Attachment 4). The concern here is not limited to 

inhalation, but also localized external exposure and exposure to unmonitored workers. Some 

consideration should be given to this route of exposure in terms of internal as well as external 

exposures. 

The University of Chicago was the first operating contractor at ORNL. As a result, many of 

those involved with work at the Metallurgical Laboratory relocated to Oak Ridge to assist in 

getting ORNL operational. For those individuals who terminated their employment prior to the 
 

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end of the University of Chicago contract, personal records were shipped back to the University 

of Chicago (Deal and Hart 1949). 

5.1.3 Air Sampling 

As discussed above, there is an absence of bioassay data for the period 1943–1947. NIOSH has 

located 1,483 air-monitoring records ranging from 1944 to 1947. Presumably, these 1,483 

ORNL specific air samples have been used to determine dose to unmonitored workers. The 

methodology presented in the TBD requires some clarification. For example, if the ORNL 

samples were used, what was the type and location of the sample in relation to the individual 

exposed? The TBD references OTIB-0018, Internal Dose Overestimates for Facilities with Air 

Sampling Programs (Brackett and Bihl 2005); however, it is unclear whether the OTIB is being 

used in the analysis. The sensitivity of survey instruments, locations of the air sampling, and air 

flow studies of the buildings were not considered. These factors would impact the accuracy of 

the air concentration data. 

In 1990, the Tiger Team evaluated all elements of the radiation protection program, including the 

air monitoring program. The following statement was made with respect to air sampler 

placement (DOE 1990): 

A thorough study of air flow patterns in ORNL facilities has been completed. This 

study indicates that the placement and air sampling number is inadequate. It will 

provide the basis for decisions on the location and type of additional air samplers 

and air monitors to meet the requirements. 

If the air samples were not appropriately placed in 1990, this leads to questions on whether they 

were appropriately placed in the 1940s. This could significantly affect the result. Several 

technical studies, including the recent 2003 Y-12 study, Practical Use of Personal Air Sampling 

(PAS) Data in the Internal Dosimetry Program at the Y-12 National Security Complex (Snapp 

2003), and Nuclear Regulatory Commission’s NUREG-1400, Air Sampling in the Workplace 

(Hickey et al. 1993), demonstrate that using air concentration data would lead to underestimating 

the worker intakes and subsequently the internal exposures. 

With the unknowns and shortcomings associated with the use of airborne concentration data for 

estimating missed dose, further evaluation should be provided to make sure this approach is 

bounding for unmonitored acute and chronic intakes. 

5.2 ISSUE 2: INADEQUATE CONSIDERATION OF MISSED DOSE FROM OTHER 

RADIONUCLIDES 

Although the Occupational Internal Dose TBD, ORAUT-TKBS-0012-5 (Bollenbacher et al. 

2006), acknowledges their existence, the TBD does not adequately address potential doses from 

secondary or so-called “exotic” radionuclides. The focus of the TBD is on “radionuclides likely 

to produce a measurable internal dose,” including uranium, activation products, fission products, 

and transuranics. Numerous radionuclides were handled at ORNL ranging in quantities from 
 

 

 

 

 

 

 

 

 

 

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fractions of a gram to kilograms. The amounts of particular radionuclides handled changed over 

time and has changed with ORNL’s mission. The basis of dose assignment for monitored 

workers is based on in-vitro bioassay data. For unmonitored workers with the potential for 

exposure, a co-worker dose is assigned. Radionuclides for which co-worker dose was assigned 

included strontium, uranium, plutonium, Am-241, Cs-137, Ce-144, and Ru-106 (Kennedy 2005). 

Although the TBD presents MDAs for a number of radionuclides, it is unclear what this data is 

used for, and whether it is used to estimate dose. 

Both R&D and production operations have contributed to the extensive and diverse list of 

radionuclides historically present at ORNL. Reactors have been used to irradiate elements. The 

long-term operations and R&D activities at ORNL provided constant opportunities for workers 

to come in contact with radionuclides spanning the periodic table. Table 5.1 shows the 

radionuclides handled by ORNL that have not been considered in the missed dose calculations or 

co-worker analyses. The radionuclides have been limited to those with a half-life greater than 

1 day. Those radionuclides with half-lives less than 1 day have been excluded. The first date the 

material was identified in the reviewed literature and the year from the earliest bioassay have 

been provided. 

Table 5-1: Radionuclides Handled at ORNL without Consideration in the Technical Basis 

Document 

Radionuclide Predominant 

Decay Mode Half-Life Radionuclide Predominant Decay Mode Half-life 

Am-243 Alpha 7,370 y Mn-52 EC/Gamma 5.591 d 

As-74 Beta/Gamma 17.77 d Mn-54 EC/Gamma 312.12 d 

As-76 Beta/Gamma 1.07788 d Mo-99 Beta/Gamma 65.94 h 

As-77 Beta/Gamma 38.83 h Na-24 Beta/Gamma 14.959 h 

Au-198 Beta/Gamma 2.6951 d Nb-95 Beta/Gamma 34.9749 d 

Ba/La-140 Beta/Gamma 12.752 d/1.6781d Ni-63 Beta 100.1 y 

Ba-133 EC 10.52 y Ni-65 Beta/Gamma 2.5171 h 

Bi-207 EC 31.5499 y Np-234 EC/Gamma 4.4 d 

Bi-210 Beta, no gamma 5.013 d Os-185 EC/Gamma 93.5999 d 

Bk-249 Beta, no gamma 320 d Os-191 Beta/Gamma 15.4 d 

Br-82 Beta/Gamma 35.2999 h Os-193 Beta/Gamma 30.11 h 

C-14 Beta 5,730 y P-32 Beta 14.26 d 

Ca-41 EC 103,000 y P-33 Beta 25.34 d 

Ca-45 Beta 162.61 d Pa-231 Alpha/Gamma 32,760 y 

Cd-109 EC 461.3999 d Pa-233 Beta/Gamma 26.9669 d 

Cd-115 Beta/Gamma 53.4599 h Pb-210 Beta/Gamma 22.2999 y 

Cf-249 Alpha 500 y Pd-103 EC/Gamma 16.9909 d 

Cf-252 Alpha 2.2 y Pd-107 Beta 6,500,000 y 

Cl-36 Beta/Gamma 301,000 y Ra-228 Beta/Gamma 5.75 y 

Cm-242 Alpha 162.8 d Rb-86 Beta/Gamma 18.631 d 

Cm-246 Alpha 4,000 y Re-186 Beta/Gamma 3.7183 d 

Co-56/57 EC/beta/gamma 77.233 d/271.7399 d Ru-103 Beta/Gamma 39.2599 d 

Co-58 EC/beta/gamma 70.86 d S-35 Beta 87.2599 d 

Co-60 Beta/gamma 5.2714 y Sb-124 Beta/Gamma 60.2 d 
 

 

 

 

 

 

 

 

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Table 5-1: Radionuclides Handled at ORNL without Consideration in the Technical Basis 

Document 

Radionuclide Predominant 

Decay Mode Half-Life Radionuclide Predominant Decay Mode Half-life 

Cr-51 EC/Gamma 27.7024 d Sb-125 Beta/Gamma 2.7585 y 

Cs-134 Beta/Gamma 2.0648 y Sc-46 Beta/Gamma 83.8099 d 

Cu-67 Beta/Gamma 61.83 h Se-75 EC/Gamma 119.79 d 

Dy/Ho-166 Beta/Gamma 81.5999 h/26.7999 h Se-79 Beta 1,100,000 y 

Es-253 Alpha 20.4699 d Sm-151 Beta/Gamma 90 y 

Es-254 Alpha 275.7 d Sn-113 EC/Gamma 115.09 d 

Eu-152 Beta/Alpha 13.537 y Ta-182 Beta/Gamma 114.43 d 

Eu-152 Beta/Alpha 9.3115 h Tc-99 Beta 211,100 y 

Eu-154 Beta/Alpha 8.593 y Th-228 Alpha/Gamma 1.919 y 

Eu-155 Beta, no gamma 4.761 y Th-229 Alpha/Gamma 7,340 y 

Fe-55 EC 2.73 y Th-230 Alpha/Gamma 75,380 y 

Fe-59 Beta/Gamma 44.5029 d Tl-201 EC/Gamma 72.912 h 

Ga-67 EC/Gamma 3.2612 d Tl-204 Beta/Alpha 3.78 y 

Gd-153 EC/Gamma 240.3999 d Tm-170 Beta/Alpha 128.6 d 

Ge-68 EC 270.82 d W-185 Beta/Gamma 75.0999 d 

Hf-181 Beta/Gamma 42.3899 d W-188 Beta/Gamma 69.4 d 

Hg-203 Beta/Gamma 46.612 d Xe-133 Beta/Gamma 5.243 d 

I-125 EC/Gamma 59.4 d Y-88 EC/Positron/Gamma 106.65 d 

I-129 Beta/Gamma 1.57 E+7 y Y-90 Beta 64.0999 h 

Ir-192 Beta/Alpha 73.827 d Zn-65 EC/Gamma/Positron 244.2599 d 

K-42 Beta/Gamma 12.36 h Zr-93 Beta 1,530,000 y 

Kr-85 Beta/Gamma 10.756 y Zr-95 Beta/Gamma 64.0199 y 

EC = Electron Capture 

The main campus of ORNL housed two major accelerators with several smaller accelerators 

distributed throughout divisions (e.g., Van de Graaff Accelerator). The Oak Ridge Electron 

Linear Accelerator (ORELA) facility consisted of a 180-MeV electron accelerator, neutronproducing targets, and evacuated flight tubes. The Oak Ridge Isochronous Cyclotron (ORIC) 

began operation in 1962. It originally supported a program of light-ion and heavy-ion nuclear 

physics until 1980. In 1980, this accelerator began operation as a beam injection for a 25-MV 

tandem electrostatic accelerator. The Holified Heavy Ion Research Facility (HHIRF), later 

known as the Holified Radioactive Ion Beam Facility, produces beams of short-lived nuclei for 

studies of exotic nuclei and astrophysics. These accelerators were involved in producing or 

analyzing radionuclides of all types (e.g., Ge-62, F-18). In some cases, the radionuclide was 

discovered at the accelerator (e.g., Xe-109, Te-105). The High Flux Isotope Reactor (HFIR) 

produced Ir-92, Co-60, californium-252, and other transuranium isotopes for research, industrial, 

and medical applications (Johnson and Schaffer 1994; www.phy.ornl.gov, Attachment 2). The 

isotope production areas were involved with the processing of secondary radionuclides, 

including the following (Fleming 2006b, pp. 33–38): 

• I-131 processing facility and the separation facility for Pm-147 

• Xe-133 facility to produce Xe-133, I-131, and Mo-99 
 

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• 	 Source fabrication (e.g., curium, Co-60, Sr-90) 

• 	 Isotope separations 

• 	 Radioactive lanthanum separation 

Several ORNL operations were conducted at the Y-12 Plant. The operations at Y-12 included 

the following (Fleming 2006b, pp. 40–42): 

• 	 R&D by the Biology Division (Buildings 9207, 9208, 9210, 9211, 9220, 9224, 9743, 

9767-3, 9767-5, and 9982) 

• 	 Stable and radioactive isotope separations at the Calutron (Building 9204-3) 

• 	 Critical Experiments Facility (CEF) (Building 9213) 

• 	 Fusion Research Division (Building 9201-2) 

• 	 Engineering Technology Division (Buildings 9201-3, 9204-1) 

• 	 Cyclotron Operations (86-inch, 22-inch, 63-inch) (Building 9201-2) 

• 	 5 MeV Van de Graaff Accelerator (Building 9202-2). 

A substantial variety of radionuclides was produced or handled in the Y-12 Plant 

calutron/cyclotron area. The 86-inch cyclotron was used to investigate the radiation effects on 

material, to produce radio-isotopes, and for experiments in basic physics. Operations with this 

particular cyclotron started in 1950 (Livingston and Boch 1952, pp. 7–8): 

The 86-inch cyclotron was constructed as a fixed-frequency cyclotron to produce 

25 MeV protons. The first proton beam was observed from the cyclotron on 

November 11, 1950. Continued development has resulted in improvements of the 

stability of operations and in larger proton currents. 

The 86-inch cyclotron was involved in angular-distribution measurements of neutrons from (p,n) 

reactions in targets; angular distribution measurements of fission fragments from proton-induced 

fission of U-233, U-235, U-238, Th-230, and Th-232; measurement of alpha-particle angular 

distributions from (proton, alpha) reactors; studies of the neutron-deficient isotopes of terbium; 

and completion of an extensive program of the measurement of activation cross sections. The 

compilation of the available data shows that isotopes produced included Be-7, Ba-133, Bi-210, 

F-18, Na-22, Mn-52, Mn-54, Fe-55, Co-56, Co-57, Zn-65, Ga-67, As-73, As-74, Y-88, V-49, 

Cr-51, Ge-68, Rb-84, Sr-85, Mo-93, Pd-103, I-123, Ce-135, Cs-137, Tm-165, W-181, Au-195, 

Pb-203, and Po-208 (Livingston 1954, Gillette et al. 1962). The Calutron at Y-12 Plant was used 

to separate stable isotopes as well as radioactive isotopes. Heavy elements separated at the 

Calutron include uranium, plutonium, neptunium, and americium (Gillette et al. 1962). 

A discussion of exposure to or problems with radon or radon daughters was not included in the 

TBD. This has been a problem or issue at most DOE sites, and is noted as a concern in early 

documents at ORNL (Ferry 1944, a series of Memoranda to Col. Stafford L. Warren, March– 

June 1944) but is not addressed in the ORAUT-TKBS-0012-5 TBD (Bollenbacher et al. 2006). 
 

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This Memorandum to Col. Warren, dated March 2, 1944 (Ferry 1944), points out that air 

samples in warehouses where ore was stored contained Mz (Radon) concentrations as high as 

6,700 microcuries/liter, which was 67.0 times tolerance. Col. Warren responded in a letter dated 

March 21, 1944, with 15 points regarding the issue and recommendations (Warren 1944). Upon 

Col. Warren’s recommendations, a large number of exhaust fans were introduced into the 

warehouses to circulate and exhaust the radon concentrations. A letter to Col. Warren, dated 

June 13, 1944 (Tybout 1944), points out that even after fans were installed, one warehouse 

sample still was 26 times tolerance. It is a well-established fact that radon has been a problem 

even in enclosed areas without ore storage as a result of background buildup from natural soils 

and building materials. SC&A believes that the problems radon presents in monitoring and 

potential exposure should be addressed in the TBD. 

There are over 2,000 radioanalytical results given in Table 5-1 for tritium. SC&A finds no 

discussion about whether any of the tritium at ORNL was in tritide form. In reviewing the 

ORNL Internal Dosimetry Program Technical Basis Document, Revision 7, December 21, 2005 

(McLaughlin 2005), it appears that all tritium was assumed to be in elemental hydrogen or water 

vapor forms, HTO and T2O. Two documents reviewed, Methods and Procedures For Internal 

Radiation Dosimetry at ORNL, January 1, 1969 (Gupton 1969), and Recommended Procedures 

for Tritium, August 1958 (Nielson 1958), also assumed these forms of tritium. None of the 

reviewed ORNL TBD documents appear to even make mention of possible tritides or metal 

tritide forms. It is SC&A’s belief that this should be addressed as to whether there were no 

tritide or metal tritide forms present, or if only elemental and water forms of tritium existed at 

ORNL. Many other DOE sites have had tritides present. It should be clarified whether or not 

ORNL had such metal tritide exposure potential. This could influence the dose reconstruction 

process. 

The Occupational Internal Dose TBD, ORAUT-TKBS-0012-5 (Bollenbacher et al. 2006), does 

not appear to address the issue of missed dose from americium. When plutonium mixtures are 

present, the absence of americium above detection levels should not preclude the calculation of 

dose from americium contribution to the mixture. Table 5-10 (Bollenbacher et al. 2006, pg. 17) 

includes Am-241 in the sequential analysis. However, americium should not be inferred as being 

absent without the presence of plutonium. SC&A does not see reference to the ingrowth of 

Am-241 into these mixtures and the issues it presents in in-vivo and in-vitro monitoring for 

internal dose reconstruction. The TBD seems to be silent with regard to this issue. 

Section 5.2.3.1 on page 18 of the TBD (Bollenbacher et al. 2006) notes the following: 

...monitoring of transplutonium elements was unable to differentiate between 

such nuclides as 241Am and 244Cm. The default radionuclide to use with 

measurements involving trivalent alpha actinides would be 241Am. The detection 

sensitivity of transplutonium analysis technique is not well documented for 

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Considering the fact that Cm-244 was handled as a radionuclide, this approach would appear to 

be faulty and could lead to missed dose from curium, unless it can be proven that workers had no 

opportunity for exposure to both radionuclides. 

These potential sources of exposure from secondary radionuclides are based on a limited SC&A 

review of ORNL operational history and available health physics files. Further research by 

NIOSH into these non-traditional radionuclides should be completed in any revision of the 

ORNL Site Profile to assure that their significance and dose contribution is fully addressed. The 

TBD should include when to assign dose from secondary radionuclide uptakes, and what default 

assumptions should be provided. Bioassay techniques should be evaluated for entire periods of 

potential exposure for effectiveness in detecting other radionuclides. Appropriate methods for 

internal monitoring were not always available for all years of potential exposure. Although the 

TBD (Bollenbacher et al. 2006) acknowledges the existence of non-traditional radionuclides, 

adequate direction is not provided on how and when to assess potential missed dose for them. 

The site profile is reliant on the existence of bioassay data or claimant interviews to ascertain 

potential exposure to secondary radionuclides. This puts the former worker, and particularly the 

survivor, at a disadvantage because of the lack of hazard identification and considerable secrecy. 

The TBD has not considered the possibility that due to secrecy, the energy employee may not 

have known what radionuclides were in the workplace. NIOSH/ORAU should base missed dose 

assignments on available data, technical reports, and other sources of information to ascertain 

potential exposure to secondary radionuclides. 

5.3 ISSUE 3: PROBLEMS WITH NEUTRON DOSES 

In view of several statements made in the Occupational External Dose TBD (Burns and 

Mohrbacher 2004), the use of NTA film to monitor neutron doses at ORNL raises several areas 

of concern. For example, page 23 of the TBD mentions that neutron energy spectra and neutron 

exposure data before the late 1980s is sparse, and that information is particularly lacking for 

many of the reactors that operated at ORNL early in its history. If n-p values are used instead of 

NTA dose records, these concerns are still valid, because using n-p values depends on a detailed 

knowledge of the gamma and neutron doses and neutron energy spectra at each work location as 

a function of time. 

5.3.1 NTA Film Neutron Energy Threshold 

The TBD acknowledges that NTA film would miss most neutrons below about 1 MeV and miss 

all neutrons below 0.5 MeV. The TBD also provides the fraction of dose above 0.5 MeV as a 

function of facility in Table 6F (pg. 82). However, it is not obvious that the dose reconstructor 

has sufficient detailed correction factor information and instructions available in the TBD to 

correct for the unmonitored neutron doses resulting from neutrons with less than 1 MeV of 

energy at the numerous facilities that produced neutron exposures through the years. The TBD 

does not provide information on how neutron dose reconstruction for facilities with all neutrons 

below the 500 keV cutoff will be handled at such facilities as the HFIR, the Graphite Reactor, 

the Radiochemical Engineering Development Center (REDC), and the Transuranium 

Decontamination Facility (TDF). 
 

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5.3.2 Neutron-to-Photon Ratios 

Neutron-photon values are used to determine neutron doses instead of NTA film results. The 

validity of this process is highly dependent on the reliability of the photon dose data. This is 

especially important because the total dose is dependent on only one dose measurement, and 

there is no ability to cross-check the gamma and neutron doses. Additionally, at some time, the 

n-p value has to have been measured for each exposure situation and knowledge of the neutron 

energy spectra obtained. The Occupational External Dose TBD (Burns and Mohrbacher 2004, 

pg. 24) states that most neutron energy spectra, doses, and n-p values to be used in dose 

reconstruction were measured since 1989. There is no spectral data presented for the early years 

when NTA film was in use. The TBD also mentions that conditions at reactor, accelerator, and 

calibration facilities are not likely to change. This may be true for a power or even a production 

reactor, but may not be generally true for research reactors and accelerators. Values of n-p and 

knowledge of the associated neutron energy spectra used in dose reconstruction should originate 

from the working conditions during exposure, not from measurements that took place much later 

in time. Dose parameters measured 30–50 years after the doses were received may not be 

directly applicable to dose reconstruction. In light of these concerns, it would be relevant to 

address the following: 

• 	 Provide information on early n-p values and neutron spectra available to be used for 

neutron dose reconstruction. 

• 	 Provide information concerning differences and similarities between early and later n-p 

values and neutron spectra investigated to date. 

• 	 If early n-p values and neutron energies spectra are lacking, present justification for using 

later data. 

5.3.3 Neutron Quality Factors 

Several items in the Occupational External Dose TBD (Burns and Mohrbacher 2004) concerning 

neutron dose adjustments need clarification: 

• 	 The TBD (page 68) instructs the dose reconstructor to correct the recorded neutron dose 

to account for the difference between the neutron quality factors in NCRP-38 (NCRP 

1971) and ICRP-60 (ICRP 1990). However, on page 69, it again recommends applying a 

factor of 2. This could confuse the dose reconstructor and the dose may get doubled 

twice. 

• 	 Attachment 6C, page 79, of the TBD lists n-p values measured during 1989–1991 in 

different facilities. The footnote indicates that the dose reconstructor should double the 

geometric means, minimum, and maximum to account for ICRP-60 radiation weighting 

factors. However, only neutron-to-gamma dose ratios are listed, not geometric means, 

etc. It is unclear if the dose reconstructor is to double the n-p values listed in this table 

when performing neutron dose reconstruction. 
 

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These items could lead to increasing the neutron dose more than is technically justified. 

5.3.4 NTA Film Fading 

The TBD makes no mention of NTA film track fading. Track fading can result in sufficient loss 

of dose data, especially for tracks resulting from lower-energy neutrons and/or in high humidity 

environments. Track fading from lower-energy neutrons has been observed as high as 56% per 

2-week exchange at Mound Laboratory (Proctor and Alguitifan 2004). This area needs to be 

addressed to determine if the exposure conditions at ORNL warrant track-fading correction 

factors for NTA film when used for neutron dose reconstruction or in determining n-p values. 

5.4 INFORMATION NEEDED CONCERNING EARLY DAY DOSE DATA AND 

SELECTION OF WORKERS FOR BADGING. 

5.4.1 Dose Data for the Earlier Years 

The Occupational External Dose TBD (Burns and Mohrbacher 2004) outlines the general 

beta/photon dosimetry techniques and dose reconstruction procedures for skin and penetrating 

doses. Attachment A of the TBD, pages 76–77, provides the recommended beta/gamma energy 

groups for the IREP program as a function of major work location at ORNL. Neutron 

adjustment factors are summarized in Attachments 6B–6F. However, the TBD only provides 

general information concerning worker monitoring and dose data. The TBD states on page 17 

that some workers were badged with experimental film before June 1944, but that these records 

may not be reliable. Therefore, it would appear that there is not sufficient beta/photon dose data 

to allow dose reconstruction for workers during 1943–June 1944. In order to evaluate the 

sufficiency of dose data for these early years (1943–1960s), the TBD needs to provide the 

following: 

(1) Information concerning: 

• The total number of workers each year 

• The number of workers monitored (and recorded) each year 

(2) The number of workers monitored each year for: 

• Beta/low-energy photon 

• Photons 

• Neutron 

(3) Details concerning the doses recorded each year, such as: 

• Minimum and maximum doses 

• Median dose 

• Percent of zeros and LOD recorded 

• Meaning of blank entries 

(4) Handling of abnormal readings: 

• Black/unreadable films 

• Lost, unreturned, damaged, contaminated badges 

• Unexpectedly high readings 
 

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5.4.2 Selection of Workers for Badging 

The Occupational External Dose TBD (Burns and Mohrbacher 2004) does not provide sufficient 

details concerning who was badged, and when, to ensure that workers were sufficiently 

monitored to allow for technically sound dose reconstruction. Table 6-2, page 16, of the TBD 

provides a list of the characteristics of dosimeters from 1944 to the present worn by radiation 

workers at ORNL, but does not describe what defined a radiation worker. The TBD (pp. 14–15) 

provides information on types, classes, color/dots, and usage of dosimeters, but no details 

concerning who was considered a radiation worker. Therefore, it is not known if the appropriate 

workers were monitored. On page 11 of the TBD, it states that NTA films were only read if the 

health physicist recommended it to be processed; this policy could lead to a false sense of being 

monitored, but the neutron dose went unrecorded on the worker’s dose record. The TBD does 

not provide sufficient details to ensure that the workers that needed to be monitored were 

actually monitored and their doses recorded for claimant-favorable dose reconstruction. To 

evaluate the sufficiency of monitoring the TBD needs to address the following: 

• 	 Provide a definition of a radiation worker as used at ORNL 

• 	 Investigate what policies were used over the years to determine what workers were 

badged 

• 	 Determine how, and to what extent, these policies were enforced 

• 	 Determine if there were groups that may have been badged after the fact (i.e., the 

radiation hazard may have gone unrecognized in the beginning), especially in the early 

years 

• 	 Determine if the maximum exposed were badged, or if it was cohort/rotational/random 

badging 

• 	 If it is postulated that the maximum exposed were badged, provide information/data to 

support this fact 

5.5 ISSUE 5: LACK OF DOSE ASSIGNMENT PROCEDURE FOR UNMONITORED 

WORKERS 

The Occupational External Dose TBD (Burns and Mohrbacher 2004) does not provide a defined 

procedure to assign dose to unmonitored workers. Section 6.5.1 briefly mentions LODs, and 

provides a listing of the LOD and exchange frequency as a function of time (Table 6-24, pg. 69). 

However, this should only be applied to the dose missed by the dosimeter worn by a worker, not 

the dose missed because a worker was not badged. This applies to neutron as well as photon and 

beta doses. During these early years, an unmonitored worker could have received dose without 

management or the worker being fully aware of the hazards. The TBD needs to provide 

technically sound dose reconstruction procedures for assigning doses to unmonitored workers, 

especially in the early years (1943–1960s) when radiation hazards were not always recognized or 

effectively addressed. OTIB-0021 (Merwin 2004) was issued on December 29, 2004, after the 

Occupational External Dose TBD (Burns and Mohrbacher 2004, issued on August 11, 2004). 

OTIB-0021 contains some recommendations for maximum/minimum (not best-estimate) beta 
 

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and photon (not neutron) doses for 1943–1985. It is uncertain how co-worker data in Table 2 of 

OTIB-0021 can contain ORNL co-worker data for 1943 and 1944. The TBD seems to contradict 

this when it states on page 17 that film dosimeter readings (results) encountered in ORNL 

personnel monitoring histories before June 25, 1944, could be unreliable, because the use of 

dosimeters during this time was experimental. These two documents appear to be in conflict 

concerning this issue. 

5.6 ISSUE 6: LACK OF DATA VALIDATION AND VERIFICATION 

The validation and verification of the data used in dose reconstruction has not been adequately 

completed. There are indications that additional bioassay data exist that are not reflected in the 

database obtained by ORAU for the calculation of MDAs. For example, we became aware that 

ORNL has not fully consolidated all the occupational exposure records, indicating that some 

records may not be complete. Also, the completeness and accuracy of the external dosimetry 

data may require further verification to ensure field-recorded dose results were integrated into 

occupational exposure records (OERs). This adds to uncertainty of these data. Finally, the 

environmental air sampling data ratios used in the development of co-worker dose from Ru-106, 

Ce-144, and Cs-137 should be further justified. 

5.6.1 Data Completeness 

OTIB-0034 (Kennedy 2005) indicates that the co-worker doses were derived from a dosimetry 

database used by ORISE in epidemiologic studies: 

Bioassay results for ORNL were obtained from the Oak Ridge Institute for 

Science and Education (ORISE) Center for Epidemiologic Research (CER) 

Dosimetry Database, which contains records from ORNL site for the period 

1951–1988. ORISE obtained the database for the purpose of conducting an 

epidemiology study of site workers. The database results are in units of 

disintegrations per minute (dpm)/24 hours (Kennedy 2005, pg. 6). 

The MDA values were derived from what appears to be a second database. The Internal Dose 

TBD indicates this database contains results from 1947 to 1988 (Bollenbacher et al. 2006, 

pg. 34). 

The electronic data that were provided by ORNL for use in estimating isotopic 

MDAs came from a project performed in the early to mid-1990s to convert 

hardcopy data over to dBase IV database. Funding ran out on the conversion 

project in the mid-1990s and the entire set of data was never completely 

converted, but a significant number of results were made available for our use 

(Bollenbacher et al. 2006, pg. 9). 

SC&A is concerned that the database used for the MDA calculations may be incomplete, and 

additional pertinent bioassay data to dose reconstruction for monitored and unmonitored workers 

may exist. For example, Table 5-9 provides an MDA for tritium in urine starting in 1961. An 
 

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early procedure for body fluid analysis, dated January 1957, states the following (Brown et al. 

1957): 

Body fluid analysis are made routinely by the Bio-Assays unit for H-3, Po, Pu, Ra, 

Sr, U, Gross Alpha (Th, Pu, Am, Cm) and Gross Beta (including K-40). 

Potential for tritium exposure extends back to at least the 1950s. For example, the Homogeneous 

Reactor Experiment – II used a fuel solution of enriched uranium sulfate (U2SO4) dissolved in 

heavy water (Huang et al. 1984). This reactor was in operation from 1957–1961. Reactors using 

heavy water at the Savannah River Site (SRS) contributed significantly to tritium dose. 

Although the database was evaluated for duplicate records and usability, the database was not 

evaluated for completeness. 

To further complicate matters, it is unclear whether the hardcopy records provided are complete. 

Occupation Exposure Records (OERs) were maintained in several different locations, making it 

difficult to readily retrieve an individual’s complete exposure record. In the 1990s, the 

Laboratory began a project to consolidate OERs into a single location. This process was not 

completed, due to lack of funding. The TBD states the following (Bollenbacher et al. 2006, 

pg. 14): 

Many hardcopy records have been consolidated into individual personal records 

folders. However, this compilation is incomplete, with records for only employees 

with last names beginning with A through G. 

One source of information absent from both the medical and radiological files at ORNL are the 

records of voluntary human radiation experimentation by ORNL workers. Two such studies are 

described below: 

During 1944 and 1945 the Clinton Laboratories, in cooperation with the Medical 

Division, placed P-32 beta-emitting plaques on the arms of 18 employee 

volunteers to study the skin erythema dose (Morgan and Peterson, 1999). 

And, 

In 1962, S.R. Bernard of the ORNL HP Division administered I-131 to himself 

and other volunteers in his research group. The level of activity neared the 

maximum permissible concentration for members of the public (Morgan and 

Peterson, 1999). 

These situations can result in significant dose consequences to particular organs and should be 

included in the dose reconstruction of claimants involved. According to ORNL staff, the 

personnel involved in these activities are listed in a database. NIOSH ought to evaluate whether 

any claimants are included in this database and retrieve reports or other documentation relevant 

to these experiments. 
 

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5.6.2 Data Accuracy 

During site expert interviews, former workers raised concerns to SC&A regarding the 

completeness and accuracy of historic ORNL radiation exposure records. Particular concerns 

were related to unauthorized practices, such as removal of routine film badges during high dose 

rate jobs, wearing film badges under protective clothing versus over clothing during routine 

work, and removing film badges. These issues were allegedly a systemic issue over the site as 

being referred to by Karl Morgan in his book (Morgan and Peterson 1999): 

Careless former college professors presented disdain for health physics 

restrictions. Ironically, the trust and high respect we received from the blue 

collar workers sometimes failed to transfer to certain scientists. 

Morgan also discussed the combination of the film and security badges (Morgan and Peterson 

1999): 

Although we issued a film badge to everyone exposed to ionizing radiation, some 

employees left their badges at home or on the desk in their office....we arranged 

to combine the film base with the security badge. Now the film badges contained 

space for a picture of the employee and his or her security number. No one dared 

be seen in the restricted area without the proper badge. 

One particular situation relayed to SC&A occurred in the first half of 1981. Approximately 

70 pipefitters and 20 welders were brought from all over the site to assist with a Cell Tear Out 

Site-wide in Building 3517. The stay time for this particular job was very short, and air-supplied 

suits had to be worn because of the radiation and contamination levels. The crafts personnel 

indicate that the dosimeter was removed and they were sent in with PICs. Health Physics 

personnel indicated that the routine quarterly dosimeter was removed, and a special job-specific 

dosimeter was used. HP indicates the special dosimeter for that job clipped to their inner set of 

coveralls. Their routine quarterly dosimeters were removed while they were wearing the jobspecific dosimeters. About half of the workers were issued PICs. Some workers were also 

issued finger rings (extremity dosimeters). The special dosimeter, PIC, and finger ring readings 

were recorded on Radiation Survey Complex Reports (see Attachment 2). A review of the 

Radiation Survey Complex Report for Building 3517 for March 1981 indicated that 28 

maintenance employees had pocket meter readings in excess of 100 mR/week. Forty-four (44) 

maintenance employees had “films processed out of routine.” The non-routine badges listed for 

the 44 maintenance employees appear to have been worn for a period of 1–6 days. There was no 

mention of bioassay for these personnel on the Building 3517 form. The results listed appear to 

indicate a predominance of beta exposure (RSCR 1981). The practice of using special dosimetry 

or even pocket dosimeters, in lieu of routine film badges, should be evaluated to ensure these 

doses are reflected in the Occupational Exposure Records and associated databases. The 

increase in frequency of exchange and relative shielding of the beta radiation could also 

influence any missed dose in these situations. 
 

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SC&A is also concerned about the approach taken for the determination of co-worker dose from 

additional radionuclides (i.e., Ru-106, Cs-137, and Ce-144). OTIB-0034 (Kennedy 2005, pg. 11) 

states the following: 

To account for additional intakes, an evaluation of air monitoring data from the 

ORNL perimeter reporting stations for the period 1975 through 1984 was 

conducted. The approach was to develop the ratios of the isotopic concentration 

ratios of other radionuclides to the concentration ratio of Sr-90 report in the air 

monitoring data. 

During the Tiger Team assessment of ORNL, 7 of the 18 ambient air sampling stations were 

examined. Three of these samplers were not at the recommended height of 2.0 meters. Four of 

the air sampling stations were too close to buildings and trees. Although they only looked at 7 of 

the stations, there was concern that other stations may have deficiencies. The Tiger Team also 

found that 6 of the 11 radionuclide stack monitoring systems examined had some air monitoring 

deficiencies (DOE 1990): 

... poor sample extraction sites, unsuitable sample transport line configuration, 

absence of air flow measurements, an inappropriately designed sampling probe, 

and a non-operational strip chart. 

There are several issues with the use of ratios to determine occupational dose from Ru-106, 

Cs-137, and Ce-144. First, an environmental air sample represents a conglomeration of 

radionuclides from multiple release points, rather than being facility specific. Second, as 

indicated above, there are deficiencies with the stack monitoring and ambient air sampling 

program that have not been taken into account. Finally, it is not clear how the period of 1975 

through 1984 is representative of years outside this range, given that operations at ORNL are 

transitional. 

In summary, SC&A is concerned about the lack of verification and validation of the databases 

used to determine MDAs and co-workers’ dose. There appears to be data absent from these 

sources that is pertinent to dose reconstruction. There is also concern over the completeness of 

the hardcopy records supplied by ORNL to NIOSH. The reliance on an incomplete database and 

dispersed exposure records may raise questions regarding the completeness and accuracy of 

internal and external dose estimates. As a compensatory process, SC&A is not aware of any 

effort to collect bioassay data known to be absent from this database. SC&A is also not aware of 

any effort to independently validate its reliability by comparing the electronic bioassay results 

with available dosimetry printouts and forms, logbooks, air sampling data, or other sources. This 

type of validation and verification is recommended. 

5.7 ISSUE 7: THE TBD FAILS TO ADEQUATELY DEFINE AND ASSESS 

OCCUPATIONAL MEDICAL EXPOSURE 

The current medical exposure and dose guidelines, as presented in (Kathren 2003), goes a long 

way in assuring that all occupational medical exposures are reasonably included in determining 
 

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the overall dose estimations for claimants. Unfortunately, the interpretation, to date, by the 

contractor (ORAU) has not been applied conservatively to be claimant favorable. The 

Occupational Medical Dose TBD (Fleming 2006a) assumes an interpretation, which has been 

also considered and applied at other sites, such as the Mound Plant and Los Alamos National 

Laboratory (LANL), Paducah, and Pinellas. To this extent, the assumption that medical 

procedures are limited to only one pre-employment chest x-ray and chest x-rays that are part of 

routine physical exams, may substantially underestimate worker medical exposure, when 

evaluating occupational medical exposure. 

In more recent documentation (OTIB-0006, Revision 3, Kathren and Shockley 2005), it is 

concluded that other examinations should be included, such as special screening exams (e.g., 

respiratory protection, beryllium workers, asbestos workers, etc.) and termination exams. The 

Occupational Medical Dose TBD (Fleming 2006a) does not recognize this change from the 

previous Revision 2 of the OTIB-0006 (Kathren 2003), and also assumes that special chest 

radiography for respirator certification, beryllium, and asbestos workers, and food handlers are 

accomplished as part of the routine physicals. This is not documented in the medical TBD. 

Another factor not discussed in the TBD is the potential and impact of x-ray procedures utilized 

by medical authorities to do special screenings that are performed outside the frequency 

suggested in the TBD or at alternate locations. The Oak Ridge Reservation had numerous sites 

and contracted with numerous radiology services and hospitals that provided these services upon 

request. 

The TBD (Fleming 2006a) makes the conclusion that chest examinations are often quite limited 

after 1970, after which chest x-rays were voluntary. Oak Ridge also did upper GI fluoroscopy 

and lumbar spines up through 1953; however, no specific recommendation is given to dose 

assessors. It is suggested the policy after 1990 called for a chest x-ray every 3 years before age 

40, and every 2 years after age 40, but nothing is documented. To the contrary, there is ample 

evidence that chest x-rays were often provided on a voluntary basis to nearly all workers, usually 

on an annual basis. The majority of workers had chest x-rays annually as a routine at DOE sites 

until the mid-1980s, when Federal guidelines warning against routine screening were first being 

enforced. 

After discussion with NIOSH personnel, it was their decision to limit occupational medical 

exposure to those chest exams described above, except for some lumbar spine exams in 1950 to 

1953, and to include all other exposure as part of worker non-occupational medical dose. SC&A 

believes such an interpretation is not claimant favorable to those most at risk. Our concern is 

that specified “high-risk” workers, those most likely exposed to radiation and beryllium, would 

be at risk of having an incomplete dose assessment if not all radiation associated to medical 

screening for job-related activities were included. Since all radiation provides some risk, and 

arguably, is cumulative, workers warrant consideration of all forms of work-related x-ray 

exposure to be claimant favorable. SC&A believes NIOSH should review its interpretation of 

included medical exposure, and should reasonably adopt a broader interpretation of occupational 

medical dose, as provided in the most recent version of OTIB-0006 (Kathren and Shockley 

2005). 
 

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5.8 ISSUE 8: TECHNIQUES AND PROTOCOLS INCREASE UNCERTAINTY OF 

DOSE CONVERSION FACTORS LISTED IN THE TBD 

Section 3.5 of the Occupational Medical Dose TBD (Fleming 2006a) fails to describe adequately 

all the information upon which to establish beam quality for x-ray units in use from 1943. In 

1947, the site documented installation of a single phase Picker R-2 unit. There is only limited 

documentation to show that the Picker unit, in use from 1947 through 1963, had added 

filtration—approximately 1.0 mm of aluminum, as first measured by Gupton, in the 1958 

surveys (Gupton 1958a and 1958b). In the absence of definitive tube output measurements, the 

TBD directs the use of default values and dose conversion factors (DCFs) derived from ICRP 

Report No. 34 (ICRP 1982). These values are then applied to determine organ doses using 

Tables A.2 through A.8 of ICRP Report No. 34 (ICRP 1982). An issue of concern is that the 

DCFs are derived using a default half-value layer of 2.5 mm Al for Type 1 units in use from 

1946–1980 (whereas the Picker unit had substantially less filtration). 

The TBD (Fleming 2006a) provides little documentation to support the assumed techniques and 

protocols applied to calculate the dose, which is mainly derived from NCRP Report 102. The 

TBD states that a PA chest x-ray was typically the only view taken. It is an undocumented 

assumption in the TBD that exams required only a PA view. SC&A has inquired whether 

definitive protocol existed to validate that chest exams possibly included PA views and LAT 

views on a limited basis. NIOSH has acknowledged in other TBD reviews that the lack of 

verifiable protocols is a generic problem at many sites, has planned to search all available 

records, and will include pertinent records and references in any future revision of this section of 

the TBD. 

The TBD is also deficient in that little documentation exists to validate x-ray protocols, 

equipment maintenance, and upkeep records. 

5.9 ISSUE 9: FREQUENCY AND TYPE OF X-RAY EXPOSURE IS UNCERTAIN. 

The Occupational Medical Dose TBD (Fleming 2006a) relies on a limited review of archived 

medical records to establish frequency assumptions. The assumption of one chest radiograph 

(PA) after 1970 on a voluntary basis is not reasonably conservative, in that workers could 

essentially request an x-ray or be subject to special screening exams. The frequency of 

screenings, and number and type of workers receiving extra annual x-rays is not evaluated. 

The TBD (Section 3) does not provide any documentation or references to support the 

assumption that only a limited group of workers received annual x-ray exams after 1970. To the 

contrary, up until about 1985, most DOE sites performed chest x-rays almost on a voluntary 

basis. DOE medical program reviews documented during the early 1990s showed many sites 

still used chest radiography as a general screening exam. Most workers accepted chest x-rays, 

even though the job did not require it. Also, the assumption that workers in special exposure 

categories, such as beryllium workers, were given chest x-rays only as part of their routine 

physical is not well-documented and not consistent with special screening guidelines. The TBD 

applies no conservative assumption to cover such exams. 
 

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The TBD (Section 3.2) states that photofluorography (PFG) units, although generally available 

up to the late 1950s at most DOE sites, were not documented as being used at the ORNL, but 

were in use at the Oak Ridge Hospital from 1943–1947. The undocumented absence of PFG 

units at ORNL clearly has significant dose implications to workers who may have been given 

much higher doses from PFG units. The PFG unit provides a dose to the worker greater by a 

factor of 5–6, more than that delivered by conventional radiography. The TBD does not provide 

documentation for the types of equipment in use at ORNL prior to 1947. SC&A believes it is 

not claimant favorable to instruct dose assessors to assume only PFG unit use from 1943–1947. 

To be fully claimant favorable, it would be appropriate to instruct dose assessors to use an annual 

dose of 3.0 rem per year for chest radiographs, in accordance with guidelines set forth (Kathren 

and Shockley 2005), until the review of medical records evidenced no further use of a PFG unit 

at any Oak Ridge site or contractor location (two Oak Ridge area hospitals and local radiology 

clinics were also used to do x-ray examinations). 

5.10 ISSUE 10: INADEQUATE CONSIDERATION OF ENVIRONMENTAL DOSE 

FROM RADIONUCLIDES OTHER THAN I-131 AND TRITIUM. 

The Occupational Environmental Dose TBD (Fleming 2006a) focuses on onsite airborne I-131 

concentrations, onsite airborne concentrations of MFPs, onsite airborne concentrations of tritium, 

and onsite exposure rate data. Reactors’ releases and waste farms data are not adequately 

considered. 

In Section 4.1, the TBD indicates the existence of site-wide monitoring data for external 

exposures dating back to early in the Laboratory’s history, but little information was available 

regarding airborne concentrations of the two principal nuclides considered for inhalation 

exposure, i.e., I-131and tritium. Regardless of data limitations for these two isotopes, we believe 

that site contamination is not limited to these two elements only. ORAUT-TKBS-0012-2 clearly 

and rightfully stated that the ORNL site has been used to test new ideas for DOE and other 

agencies since inception in 1943. Many of the earliest buildings at the site have transitioned 

through various mission objectives. Thus, all types of radioactive contaminants were expected to 

be found at this site, not only I-131 and tritium. In Section 2.1 of the TBD, a partial list of the 

reactors that were designed and, in most cases, tested at ORNL, and then either scrapped or 

further developed elsewhere, included the LITR, CEF (at the Y-12 National Nuclear Security 

Complex), BSR)/PCA, ORR, TSR, HPRR, HRE, ANP Program, the HFIR, and the MSRE. 

Section 2.1 additionally emphasized that waste radioactive material was released from early site 

operations as gaseous, liquid, and solid effluents, with little or no pretreatment. Later, methods 

were developed to capture many of the contaminants at their source and to reduce overall plant 

emissions. In some cases, this increased direct exposures to individuals in the area and created 

locations in which incidents and spills occurred (Fleming 2006). 

Radiological discharges from each stack were unique because of the variety of research activities 

on site. Emissions from ORNL typically consist of solid particle, absorbable gases, tritium, and 

nonabsorbable gases (DOE 1990). In section 2.2.7, the Site Description TBD clearly identified 

an array of isotopes that were released from only one building; the Graphite Reactor. The quote 

of that section is as follows (Fleming 2006b, pg. 10): 
 

 

 

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Radionuclides produced in the reactor included 35S, 32P, 31Si, 42K, 41-45Ca,46Sc,

51Ti, 59Fe, 55Fe, 60Co,65Ni, 64Cu, 75Se, 110Ag, 114In, 115Cd, 124Sb, 152Eu,154Eu, 155Eu,

182Ta, 185W, 185Os, 191Os, 193Os, 204Tl, 206Tl, 210Bi, 24Na, 76As, 82Br, 86Rb, 99Mo,

198Au, 131I, 141-143Ce, 14C, and 192Ir. Activation of the cooling air resulted in the 

chronic release of 41Ar from the Building 3018 stack whenever the reactor was 

operating. In 1948, 41Ar releases totaled 540 Ci d-1 at a reactor power of 

4,000 kW(t) and an exit airflow rate of 51,000 cfm. Fuel slug ruptures in 1947 

resulted in emissions of 131I, 133I, 129I, 137Cs, 90Sr, 85Kr, 133Xe, 239Pu, 103Ru, 106Ru,

144Ce, 140La, 140Ba, 95Zr, 89Sr, 235U, 238U, and 95Nb. 

In addition, Table 2-2 identified many radioactive elements of concern at the North Tank Farm 

(Building 3023 or 206). These radioisotopes of concern include U-233, U-235, Np-237, 

Am-241, Pu-238, Pu-239, Sr-90, Cs-137, Co-60, and tritium. Radioisotopes of concern from the 

South Tank Farm (Building 3507or 206) include U-233, U-235, Am-241, Pu-238, Pu-239, Sr-90, 

Co-60, and tritium (Fleming 2006b). 

The X-10 facility handled irradiated thorium, radioactive lanthanum, and plutonium. Airborne 

effluent emissions from that plant included U-233. This isotope has a much higher specific 

activity, and somewhat higher effective dose coefficients, than the other uranium isotopes. Thus, 

if this nuclide was included in the total uranium (expressed in :g/m3) measured at the Y-12 

facility, it would have a significantly greater radiological impact. Furthermore, when U-233 is 

produced by irradiating Th-232 in a reactor, U-232 is an inevitable byproduct. Although its mass 

concentration is in the range of 5–50 ppm in the uranium metal, it has an extremely high specific 

activity, due to its short (68.9-y) half-life, and has effective dose coefficients that are several 

times higher than those of U-234. 

The Occupational Environmental Dose TBD (Burns 2004b) also has not considered 

environmental exposures to ORNL personnel assigned to areas at the Y-12 Plant. Environmental 

air monitoring at the Y-12 Plant focused on evaluating releases of U-234, U-235, and U-238. It 

appears reasonable that if the Y-12 workers were exposed to airborne uranium, the ORNL 

workers located at Y-12 would also be exposed to the same uranium. 

Because of these facts, additional elaboration in the document is needed to explain the 

methodologies used to take into account the environmental contamination consequences and 

exposure of involved and uninvolved workers to this wide array of radioisotopes. 

5.11 SECONDARY ISSUES 

5.11.1 Secondary Issue 1: Potential for High-Fired Oxides 

The ORNL Site Description TBD (Fleming 2006b, pg. 25) states the following: 

In addition, Building 4508 housed the Fuel Cycle Alpha Facility (FCAF), which 

fabricated plutonium oxide and plutonium-uranium oxide fuel pellets, and was 
 

 

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used for coating sol gel-derived microspheres with pyrolytic carbon. In addition, 

special target materials for the HFIR were developed and fabricated. 

These processes occurred in Building 3019 and 4508. It is noted that gloveboxes were equipped 

to synthesize, press, and sinter pellets of Pu and U nitrides in Building 4508. It would appear 

that there may have been the potential for creation and possible exposure to high-fired oxides in 

some of these processes. The TBD makes no reference to the potential of producing such forms, 

or how they would be handled if such insoluble forms were produced in these processes. The 

possibility for or against creation of such forms should be addressed, since Super S forms have 

been found and noted at most other DOE facilities processing plutonium. 

5.11.2 Secondary Issue 2: Ingestion Pathway is Not Addressed 

The Occupational Internal Dose TBD (Bollenbacher et al. 2006) does not address or consider 

potential exposure from the ingestion pathway, especially a potential issue in the early years. 

Many of the radionuclides handled at ORNL were highly soluble. Review of the ORNL Internal 

Dosimetry Program Technical Basis Document, Revision 7, December 21, 2005 (McLaughlin 

2005), references ingestion 74 times. McLaughlin 2005 addresses ingestion intake retention 

fractions (IRF), and page 57 of that document presents a section on “Internal Dosimetry 

Software Used at ORNL,” which discusses acute ingestion and chronic ingestion pathways. 

Page 69 of the TBD addresses DCFs for ingestion intakes. It appears that the Internal Dosimetry 

Program at ORNL recognized and calculated internal doses from ingestion intakes. The 

Occupational Internal Dose TBD (Bollenbacher et al. 2006), however, appears silent on this 

issue. SC&A suggests that this pathway needs to be addressed in the TBD. 

5.11.3 Secondary Issue 3: Lack of Explanation for Determining MDAs 

Section 5.1.5 (Bollenbacher et al. 2006, pg. 12) states that ORNL has entered much historical invitro monitoring data into a database. In Section 5.1.1, it states that the values in Table 5-1 do 

not reflect the total number of in-vitro bioassays performed by ORNL in this period, because not 

all hardcopy records are in the database. These statements are difficult to interpret, and the 

numbers in the tables seem to match fairly well. However, the numbers of samples analyzed and 

used to determine historical MDAs appear to be extremely small in many cases with a sample 

size less than 10, and there are a large number of single samples used for determining the MDAs 

(Table 5-1, pg. 10, and Table 5A-1, pg. 36). Are these really most of the analyses that were done 

on workers for many of these radionuclides? NIOSH should further evaluate a method for 

determination of MDA values with an adequate number of data points including those for other 

radionuclides. 

There needs to be much more explanation of what was going on with the large variations in 

MDA values (dpm/24-hour sample) for some of the radionuclides listed in the TBD (Table 5-9, 

pg. 16, and Table 5A-2, pg. 39). Specific examples include uranium and Ru-106. The uranium 

MDA (Table 5-9, pg. 16) goes from 1.4 in years 1948 through 1950, then up to 6.3 in years 1951 

through 1963, and then back down to 1.1 and 0.09 and 0.06. The MDA for Ru-106 goes from 

0.30 in years 1951 through 1959, and then jumps up to 53. Was the MDA really lower in the 
 

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early years, and, if so, why? It is obvious that MDA values would improve and go down as 

better analytical techniques and methods are discovered. The uranium analysis data in the years 

1951 through 1963, using the jump up in MDA values, raises concern for missed dose for those 

claimants. If you really had MDA values for Ru-106 of 0.036 in 1951, why would you change to 

an analytical procedure that gives an MDA of 78.9 in 1964 (Table 5A-1, pg. 39). SC&A feels 

that further explanation is needed for some of the obvious discrepancies in MDA values over 

time. 

5.11.4 Secondary Issue 4: Other Potential Medical Exposures Have Not Been Identified 

The Occupational Medical Dose TBD (Fleming 2006a) does not address the potential use of 

radiation exposure from exposure sources other than x-ray units, in the support of medical 

diagnosis. This may involve the use of isotopes, sealed sources, etc. The TBD is also deficient 

in that it does little to catalog the number, types of x-ray equipment, frequency of use, etc., 

discussed above in Issues 2 and 3. 

The below average performance at ORNL to conduct routine and preventative maintenance 

during the 1947–1990 timeframe suggests that routine maintenance of x-ray units was not likely, 

unless performed by an unknown outside contractor. Unfortunately, no records exist to evidence 

maintenance, calibrations, etc. The lack of defined protocols and basis for approval of 

radiography procedures suggests that the use of radiography was not closely controlled. The 

Occupational Medical Dose TBD does not discuss the use of portable radiography to perform 

screenings, and the potential for exposure of medical personnel or other workers without 

dosimetry devices being utilized. This is potentially an issue for the PFG unit, which was often 

van-mounted at other sites. 

The conclusion is that the TBD does little to reasonably document the variety of medical 

occupational exposures, and the lack of documentation on the type of equipment and the 

maintenance records does little to assure that a conservative and claimant-favorable estimation of 

dose is possible. This circumstance would suggest the need to reconsider a worst-case approach 

to establishing dose. NIOSH should revisit and update, as needed, Sections 3.4 and 3.5 of the 

TBD. 

5.11.5 Secondary Issue 5: Additional Factors Contribute to Uncertainties 

The Occupational Medical Dose TBD (Fleming 2006a) does not consider dose impacts due to 

less-than-optimal use of technology, such as using screens, grids, or bucky systems. The TBD 

does not consider these elements as potential contributions to uncertainty. 

The TBD does consider the potential contribution to dose that may have resulted in less-thanoptimal use of collimation, at least prior to 1970, as stated in Section 3.5 of the TBD. 

Unresolved is the concern that the DCFs are derived from ICRP Publication 34 (ICRP 1982), and 

therefore, are not comparable in terms of beam quality, which varies from unit to unit. These 

factors can contribute greatly to the dose to the chest and other organs for the units in operation 

prior to 1990, as little or no documentation exists. NIOSH has indicated in other TBDs that it 
 

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will continue to search for other available records to better define equipment use and beam 

quality, and include it in an updated version of the TBD, as appropriate. 

Uncertainty as defined is the TBD as being due to measurement error, and variation in 

kilovoltage, tube current, timers, and the source-to-skin distance. This approach is quite similar 

to the uncertainty analyses documented in other DOE site profiles. The conclusion in this TBD 

(and others) is that an uncertainty factor of +30% should be used by dose reconstructors. SC&A 

believes the uncertainty correction factor of 2.0 (being applied at other sites) is more appropriate 

in order to ensure claimant favorability. 

SC&A agrees that the TBD conservatively estimates these essential aspects of an uncertainty 

review. Unresolved is the contribution to uncertainty in dose, due to other errors introduced by 

lack of quality controls in processing equipment and lack of adherence to established standard 

operating procedures. A reasonable estimate of these contributions to uncertainty would be an 

evaluation of retake rates, per examination type. NIOSH should revisit the potential for 

significant retake rates and evaluate its potential effect on dose as part of future revisions of this 

TBD, especially as it relates to prior to 1990, when photo timing of exposures was adopted. 

The TBD does not show that ORNL applied dose minimization principles to reduce medical 

exposures. The document also does not assess or consider the likely exposure to workers who 

are referred to offsite medical facilities for follow-up. The TBD states that review of selected 

medical records and files did not reasonably show or match expected x-ray exam frequency and 

type of exam, as shown in Tables 3.2 and 3.3. Little evidence exists to document the number of 

x-ray exams provided to the average worker or for special exposure needs. 

5.11.6 Secondary Issue 6: Average Annual Environmental Exposures 

The Occupational Environmental Dose TBD (Burns 2004b) states that a number of conservative 

approaches and assumptions had to be employed to develop the requisite concentration and 

intake data. These results should be considered more bounding than representative. SC&A 

believes that this could be a reasonable approach only if the approach is proven claimant 

favorable. Also, it is not clear how the annual "average" airborne concentrations and annual 

"average" exposure rates are bounding. Environmental doses to local workers following 

accidents and spills can be significant. Limiting the doses to airborne releases, as alluded to in 

Section 4.2 of the TBD, is certainly undermining the dose contributions from environmental 

sources. It is unclear how the average values of releases took into account the releases and 

deposition of large amounts of radioactive contaminants at the site. Several examples in the 

TBD clearly identified these releases and potential precipitation of radioactive particles (Burns 

2004b, pp. 6 and 7). Several quotes are given below. 

The exhaust air from the Graphite Reactor was discharged unfiltered until 1948 

when a significant particle contamination problem was found on the Laboratory 

grounds. The Graphite Reactor was found to be one of the principal contributors 

to this particulate contamination, so a filter house was added between the exhaust 

plenum and the stack. The filtration system, which became operational in 
 

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November 1948, consisted of parallel banks of roughing plus high-efficiency 

filters. Particulate releases from the 105 stack were largely mitigated once the 

filtration system was added. The 105 stack also ventilated gaseous effluents from 

the Low-Intensity Test Reactor (LITR) from 1949 until 1968 (Burns 2004b, pg. 6). 

...high-activity offgas streams were ventilated via the 205 stack prior to 

completion of the ORNL central offgas handling system about 1950 (Burns 2004b, 

pg. 7). 

Furthermore, 

Prior to 1950, airborne radioactive effluents from ORNL were discharged to 

stacks and vents serving individual facilities. However, the temporary nature of 

the Laboratory’s original mission resulted in it quickly outgrowing its waste 

handling measures once its status was changed to one of indefinite duration 

(Burns 2004b, pg. 7). 

In the Site Description TBD, similar descriptions were made to justify our concerns regarding 

the claimant-favorability approach in dose reconstruction (Fleming 2006, pg. 24): 

Other than operational problems (failed equipment, breached control systems, 

plugged feed lines) that occurred at times during RaLa operations, the worst 

incident occurred at about 5:00 p.m. on April 29, 1954. The incident was 

described as “the most serious accidental release of activity ever experienced in 

the history of the process” (Rupp and Witkowski 1955). The Building 706-D 

.............. When the fourth batch addition of acid was poured into the dissolver 

to initiate additional dissolution, a violent reaction occurred, forcing dissolver 

solution up the slug-loading chute and solution addition lines. Air monitors in the 

building immediately sounded an alarm, indicating elevated airborne radioactive 

material. The investigation that followed indicated that individuals in the 

building donned gas masks and evacuated the building soon. The release lasted 

from between 10 min to 2 hr before the scrubbers could recover and begin 

filtering radioiodine from the building. A letter written by the Laboratory Shift 

Superintendent to the ORNL Director indicated “all people involved in the 

incident and later in the high level decontamination work are being given the 

standard HP check including urine checks, etc.” (Stanley 1954). Radiation 

levels reached 100 R hr-1 on the third floor, but were reduced to 100 mR hr-1 by 

7:00 a.m. the next day; air sample results did not exceed the tolerance level of 

3.0E-08 μCi ml-1. A preliminary check of film badge results was indicated as 

having been conducted, but the reference did not provide confirmation. 

From the brief description above, several issues might require additional assessments. One of 

them is the uncertainty of the release duration. A release of 10 minutes to 2 hours is a wide 

range to accurately define dose. Also, the appropriateness of the use of site-wide average 

exposure rates was questioned by the authors of the TBD, particularly for cases in which a 
 

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composite exposure profile for an unmonitored individual is inconsistent with a known or 

asserted exposure history. The third concern is the acute exposure to 100,000 mrem/hour for 

2 continuous hours. 

Therefore, SC&A believes that the Occupational Environmental Dose TBD (Burns 2004b) 

should be supplemented with a detailed discussion of associated uncertainty analysis on airborne 

concentration data and exposure data. Also, we wonder if uncertainties have been quantified, 

and if they have been utilized in the environmental dose reconstruction. 

5.11.7 Secondary Issue 7: Inadvertent Ingestion and Inhalation of Resuspended Particles 

Not Considered 

Despite the fact that the Occupational Environmental Dose TBD (Burns 2004b) is continuously 

assuming that the air monitors at the site intercept and take into account resuspended particles, 

SC&A is concerned that the assumption may not be correct. Depending on the height of the air 

monitors, these particles can play an important role in dose reconstruction of unmonitored 

workers. For example, the extent of soil contamination at ORNL is documented in surveys that 

can be obtained from http://risk.lsd.ornl.gov/maps/x-10/x10_relsites.shtml. 

Another concern is the admission of DOE itself that contamination due to radioactive releases is 

large, particularly for certain processes in the early days of ORNL operation. Of particular 

interest is the following statement: 

...Specific attention is given to the production of RaLa and the fission products 

I-131 and Cs-137. RaLa production for nuclear weapons development reached 

its height during the late 1940s and early 1950s, which resulted in rarely 

monitored or restricted releases of I-131 and other short-lived fission products. 

 ORNL also manufactured I-131 and Cs-137 for therapeutic use in the private 

sector and its own research needs. The production and onsite use of these 

nuclides resulted in both deliberate and unplanned environmental releases. 

 Moreover, these nuclides were prevalent in the large quantities of liquid waste 

and airborne contaminants that have been discharged and monitored at the ORR 

since the 1940s. 

http://www.eh.doe.gov/ohre/new/findingaids/epidemiologic/oakridge3/intro.html 

SC&A is concerned that the resuspension factors of these contaminated soils, which should be 

used to derive doses to workers from inhalation or ingestion, were not fully addressed. As we 

indicated for SRS and Rocky Flats Plant (RFP) site reviews, there are a couple of methods to 

estimate inhalation exposure from resuspended radionuclides. These methods include the dustloading approach and the resuspension-factor approach. In addition, SC&A refers to 

Section 1.1.3 of this report regarding our concerns of the appropriateness of the air monitoring 

network and locations, and hence, the reliance on its data to intercept particles’ resuspension. 
 

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6.0 OVERALL ADEQUACY OF THE SITE PROFILE AS A BASIS FOR 

DOSE RECONSTRUCTION 

The SC&A procedures call for both a “vertical” assessment of a site profile for purposes of 

evaluating specific issues of adequacy and completeness, as well as a “horizontal” assessment 

pertaining to how the profile satisfies its intended purpose and scope. This section addresses the 

latter objective in a summary manner by evaluation of (1) how, and to what extent, the site 

profile satisfies the five objectives defined by the Advisory Board for ascertaining adequacy; 

(2) the usability of the site profile for its intended purpose, i.e., to provide a generalized technical 

resource for the dose reconstructor when individual dose records are unavailable; and (3) generic 

technical or policy issues that transcend any single site profile that need to be addressed by the 

Advisory Board and NIOSH. 

6.1 SATISFYING THE FIVE OBJECTIVES 

The SC&A review procedures, as approved by the Advisory Board, require that each site profile 

be evaluated against five measures of adequacy; (1) completeness of data sources, (2) technical 

accuracy, (3) adequacy of data, (4) site profile consistency, and (5) regulatory compliance. The 

SC&A review found that the NIOSH site profile for ORNL presents an adequate accounting of 

the primary internal and external issues related to predominant radionuclides, such as uranium, 

plutonium, and some fission products. The ORNL Site Profile falls short in fully characterizing 

a number of key underlying issues that are fundamental to guiding dose reconstruction. In some 

cases, these issues may impact other site profiles. Many of the issues involve lack of sufficient 

conservatism in key assumptions or estimation approaches, or incomplete site data or analyses of 

these data. Section 6.0 summarizes the key issues. A detailed evaluation of these issues is 

provided elsewhere in the report. 

6.1.1 Objective 1: Completeness of Data Sources 

The breadth of data sources used as a basis for the ORNL Site Profile is evident in the 472 

reports available for ORNL in the Site Profile Research Database. Two hundred twenty-six 

(226) reports were cited in the site profile references, while others served to provide confirmatory 

information, or were only recently retrieved. The NIOSH/ORAU team consulted health physics 

personnel with long histories at ORNL who have extensive knowledge of key dosimetry 

historical processes and personnel monitoring data. There was a meeting held with the ORNL 

Labor and Trades Council on November 8, 2004, in order to identify worker concerns and 

discuss the TBDs. This interaction has helped to provide valuable insight into site operations 

and processes. In addition, the issuance of supporting TIBs reflect the ongoing effort by NIOSH 

to continually improve guidance provided to dose reconstructors. 

However, the site profile falls short in its critical evaluation of pertinent records and purposeful 

use of site expert interviews to ascertain potential monitoring or records gaps throughout 

ORNL’s history, with the objective of determining the extent and significance of unmonitored 

worker dose. This issue is particularly acute among workers who were transient onsite, and 

between the Y-12 Plant and Bethel Valley locations. Until the 1990s, there was a reliance on the 
 

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field to determine when and what bioassay was required. Audits in later years indicated the 

inconsistencies that exist in the Radiological Control program, especially between the 

Bethel/Melton Valley location and the Y-12 Plant location. 

The validation and verification of the data used in dose reconstruction has not been adequately 

completed. There are indications that additional bioassay data exist, which are not reflected in 

the database obtained by ORAU for the calculation of MDAs. For example, ORNL has not fully 

consolidated all the occupational exposure records, indicating that some records may not be 

complete. Absence of bioassay data in the database is also evidenced by the existence of source 

term and bioassay techniques for some radionuclides prior to the initial MDA calculated in the 

TBD. Based on Comprehensive Epidemiology Data Resource (CEDR) data retrieved by ORNL, 

the urinalysis program was diverse and bioassay data was available for 63 radionuclides. Many 

of these radionuclides had less than 10 samples from the period of 1947–1988. Other 

radionuclides handled at ORNL were not available in the database, many of which were listed in 

the Site Description TBD. Further evaluation of secondary radionuclides is necessary to 

ascertain the potential missed dose to workers. Finally, the environmental air sampling data 

ratios used in the development of co-worker dose from Ru-106, Ce-144, and Cs-137 should be 

further justified. 

In terms of environmental dose, those workers permanently assigned to ORNL operations at the 

Y-12 Plant were not considered separately from those at Bethel/Melton Valley. The 

environmental dose measurements concentrated on I-131 and tritium exposure, while the 

predominant source of environmental release at the Y-12 Plant was uranium. The ORNL TBD 

authors are encouraged to review the Y-12 TBD to determine the unique hazards associated with 

this location. 

6.1.2 Objective 2: Technical Accuracy 

The Occupational Medical Dose TBD (Fleming 2004) provides little documentation to support 

the assumption that techniques and protocols applied to calculate the dose (mainly derived from 

Cardarelli et al. 2002) is accurate. NIOSH believes that when no information is readily available 

about the energy spectrum, it is reasonable to use the assumptions for DCFs presented in the 

Implementation Guide. 

The Occupational Environmental Dose TBD (Burns 2004b) appears to rely on emission and 

measurement data; however, it does not indicate the model used for calculations. The TBD 

generally discusses particle size; however, the actual particle size assumptions for assignment of 

internal dose have not been provided. No consideration has been given to the deficiencies in the 

stack and ambient air sampling systems; however, there is a heavy reliance on these systems to 

determine unmonitored worker dose. Exposures considered are limited to I-131, H-3 (starting 

1967), Kr-85, Xe-133, and MFPs (optional), while diverse radionuclides were handled and 

potentially released at the site. There has been no consideration of potential doses from the 

release of large uranium particles. Overall, SC&A believes that further investigation into 

environmental source terms and pathways is needed. 
 

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The Occupational External TBD (Burns and Mohrbacher 2004) may result in an underestimate 

of neutron dose. Neutron dose is determined from NTA film results, and is modified with a 

correction factor. Some facility-specific neutron energy bands are provided; however, in some 

facilities, the entire spectrum is essentially below the practical 1-MeV detection limits of NTA 

film used in the workers’ badges. From the information in the current Occupational External 

Dose TBD (Burns and Mohrbacher 2004), it is not obvious that the dose reconstructor has 

sufficient detailed correction factor/instructions available to correct for the unmonitored neutron 

doses resulting from neutrons with less than 1 MeV of energy at the numerous facilities at ORNL 

that produced neutron exposures through the years. 

Unlike other TBDs, there is a lack of defined procedures on how to assign missed dose to 

unmonitored workers. There needs to be clear direction on how to assign missed dose to ensure 

consistency among dose reconstructions and prevent confusion. 

Information available for the dose reconstruction in the early years is limited, inadequate, and in 

some cases, not available. External monitoring for 1943–1944 was limited to the use of PICs 

with some experimental badges worn. There is a lack of clear discussion on how these 

monitored and unmonitored doses are derived during this time period. Furthermore, the 

Occupational External Dose TBD (Burns and Mohrbacher 2004) questions the validity of these 

data. The monitoring practices for the years prior to 1951 require further investigation to 

determine if all exposed workers were monitored during this time period. In terms of internal 

exposure, there was an absence of routine internal monitoring until 1949. Early bioassay data 

were limited to plutonium and strontium, although other radionuclides were being handled even 

prior to 1949. 

The Occupational Internal Dose TBD lacks guidance on how to assign dose to radionuclides 

other than transuranics, uranium, activation products, and fission products. As indicated by site 

experts, ORNL handled almost everything on the periodic table at one point or another. There 

has been no screening presented to demonstrate that the secondary radionuclides (particularly 

accelerator- and reactor-produced) are of no dose consequence to the workers. Although ORNL 

handled uranium and radium in the early years, no consideration was given to occupational radon 

exposure. Information was not provided on the activity fractions for plutonium and thorium. 

Activity fractions for plutonium provide critical information for the assessment of dose from 

americium as an impurity. Dose from non-traditional chemical forms of radionuclides, such as 

high-fired oxides and tritides, were not considered. Finally, an adequate rationale for assumption 

of Am-241 in the case of transplutonium bioassay rather than Cm-244 was not provided. 

The internal co-worker dose assignment for Ru-106, Ce-144, and Cs-137 was derived by 

determining their ratio to strontium on environmental air samples. The equivalency of ratios on 

an environmental sample to the ratios that exist in an occupational setting has not been 

adequately justified and limits the dose reconstructor to a site-wide ratio only. 
 

 

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6.1.3 Objective 3: Adequacy of Data 

The completeness and accuracy of the external dosimetry data may require further verification to 

ensure that field-recorded dose results were integrated into OERs. The internal database 

currently used to determine the co-worker doses are incomplete in terms of the years covered 

(1947–1988). There is also an indication from ORNL records that bioassay procedures predated 

the first MDA presented in the Occupational Internal Dosimetry TBD (Bollenbacher et al. 2006). 

Data adequacy issues have been raised by ORNL workers and will require further investigations. 

The Occupational External Dose TBD (Burns and Mohrbacher 2004) outlines the general 

beta/photon dosimetry techniques and dose reconstruction procedures for skin and penetrating 

doses. The TBD only provides general information concerning worker monitoring and dose 

data. The TBD states on page 17 that some workers were badged with experimental film before 

June 1944, but that these records may not be reliable. Therefore, it would appear that there is 

insufficient beta/photon dose data to allow dose reconstruction for workers during 1943–June 

1944. 

The validation and verification of the data used in dose reconstruction has not been adequately 

completed. There are indications that additional bioassay data exist, which are not reflected in 

the database obtained by ORAU for the calculation of MDAs. For example, we became aware 

that ORNL has not fully consolidated all the occupational exposure records, indicating that some 

records may not be complete. Also, the completeness and accuracy of the external dosimetry 

data may require further verification to ensure that field-recorded dose results were integrated 

into the OERs. 

In summary, SC&A is concerned about the lack of verification and validation of the databases 

used to determine MDAs and co-worker dose. There appears to be data absent from these 

sources that is pertinent to dose reconstruction. There is also concern over the completeness of 

the hardcopy records supplied by ORNL to NIOSH. The reliance on an incomplete database and 

dispersed exposure records may raise questions regarding the completeness and accuracy of 

internal and external dose estimates. As a compensatory process, SC&A is not aware of any 

effort to collect bioassay data known to be absent from this database. SC&A is also not aware of 

any effort to independently validate its reliability by comparing the electronic bioassay results 

with available dosimetry printouts and forms, logbooks, air sampling data, or other sources. This 

type of validation and verification is recommended. 

6.1.4 Objective 4: Consistency Among Site Profiles 

An extensive comparison was performed by SC&A to compare and contrast the methodologies 

used in the ORNL site profile and other site profiles reviewed to date. These comparisons focus 

on the methodologies and assumptions associated with dose assessments and the derivation of 

values used to obtain a POC for individual claimants. A detailed analysis is provided in 

Attachment 5 to this report. The site description provides a comprehensive evaluation of 

activities that occurred at the different technical areas, and some of the potential hazards 

associated with these operations. This valuable data is not carried through to the other TBDs. 
 

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Overall, the default values assigned for determining medical exposure are relatively consistent 

among site profiles. The site profiles applie OTIB-0006, Revision 2, dated December 29, 2003, 

although a subsequent version was released on December 21, 2005. This version predates the 

issuance of Rev. 01 PC-1 of the TBD dated July 21, 2006. The later version of OTIB-0006 

(Kathren and Shockley 2005) provides guidance on the assignment of dose from lumbar spin 

exams, which were given at ORNL. This should be corrected in subsequent revisions of TBDs. 

Other deviations from the standard assumptions are based on site specific information. 

The ORNL Occupational Environmental Dose TBD assumes that average annual site-wide 

values are assigned to unmonitored workers. For LANL and SRS, the site-wide maximum 

ambient radiation dose is assigned when the worker location is unknown (Cehn and McDowellBoyer 2004; Scalsky 2005). The application of multiple receptor points would be consistent 

with the approach in the Hanford, SRS, and Idaho National Engineering and Environmental 

Laboratory (INEEL) TBDs. The use of measured air concentration data is consistent with the 

Y-12 Occupational Environmental Dose TBD (Ijaz and Adler 2004). 

With the LANL Occupational Environmental Dose TBD, a screening method was applied to the 

source term to determine relative dose consequences for different radionuclides. With the 

similarities in the breadth of radionuclides handled at ORNL and LANL, this screening method 

or equivalent is considered prudent at ORNL. 

Review of site profiles to date indicates that the NIOSH/ORAU team has not come to a 

consensus on what components should be considered in the environmental dose. In the case of 

the ORNL Occupational Environmental Dose TBD, no consideration was given to dose from 

contaminated soil resuspension. It is noted that other site profiles have considered radionuclides 

that were present at ORNL, but not considered in the ORNL Occupational Environmental Dose 

TBD. The Hanford TBD included plutonium, noble gases, and MFPs (Savignac 2003). The 

SRS TBD included tritium, noble gases, plutonium, and uranium (Scalsky 2005). The LANL 

TBD included tritium, noble gases, plutonium, americium, uranium, and thorium (Cehn and 

McDowell-Boyer 2004). Several of the ORNL workers were stationed at the Y-12 Plant 

location, indicating these radionuclides may not be applicable. The Y-12 Plant handled 

primarily uranium, with some processing of thorium and U-233 (SC&A 2005c). These 

radionuclides have not been considered in the assignment of dose to workers at the Y-12 facility. 

There is mention of the presence of the release radon from stacks and storage areas; however, no 

discussion was included for potential internal dose to radon. It would seem reasonable that radon 

would be present in the work place if it were being released to the environment. Radium-226 

and ores were also handled at ORNL. Radon was specifically mentioned in the Mallinckrodt 

Chemical Works (MCW) and Fernald TBDs, where high concentrations of Ra-226 were handled. 

The ORNL Site Profile fails to consider radon exposure. The impact of radon on dose 

reconstruction should be evaluated, particularly for the early years. 

During the Y-12 SEC petition review, considerable attention was given to radionuclides 

generated by the ORNL Isotope Production Group and potential exposures to Y-12 workers from 

these activities. ORNL employees were stationed at the Y-12 Plant and actively ran these 
 

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operations. The radionuclides of concern for the Y-12 Plant are H-3, Sr-90, Tc-99, Th-228, 

Th-232, U-232, U-233, U-236, Pu-238, Pu-239, Pu-241, Np-237, Am-241, Co-60, Zr/Nb-95, and 

Po-208. This is a somewhat different mixture of radionuclides than those considered for missed 

dose at the ORNL main site. The Occupational Internal Dose TBD (Bollenbacher et al. 2006) 

does not include a discussion how these are handled in dose reconstruction in terms of missed 

dose. 

6.1.5 Objective 5: Regulatory Compliance 

SC&A reviewed the site profile with respect to Objective 5, which requires SC&A to evaluate 

the degree to which the site profile complies with stated policy and directives contained in 

42 CFR Part 82. In addition, SC&A evaluated the TBDs for adherence to general quality 

assurance policies and procedures utilized for the performance of dose reconstructions. NIOSH 

has complied with the hierarchy of data required under 42 CFR Part 82 and its implementation 

guides. As mentioned above, quality assurance with respect to claimant-specific information is 

lacking, and further consideration should be given to evaluating records provided by sites, and 

how the requests for these records are communicated to the sites. In essence, if something is not 

explicitly requested, it will not be provided. 

6.2 USABILITY OF SITE PROFILE FOR INTENDED PURPOSES 

SC&A has identified seven criteria that reflect the intent of the EEOICPA and the regulatory 

requirements of 42 CFR Part 82 for dose reconstruction. Because the purpose of a site profile is 

to support the dose reconstruction process, it is critical that the site profile assumptions, analytic 

approaches, and procedural directions be clear, accurate, complete, and auditable (i.e., 

sufficiently documented). SC&A used the following seven objectives to guide its review of the 

ORNL Site Profile TBDs to determine whether they meets these criteria: 

Objective 1 − Determine the degree to which procedures support a process that is expeditious 

and timely for dose reconstruction 

Objective 2 − Determine whether procedures provide adequate guidance to be efficient in select 

instances where a more detailed approach to dose reconstruction would not affect the outcome 

Objective 3 − Assess the extent to which procedures account for all potential exposures and 

ensure that resultant doses are complete and are based on adequate data 

Objective 4 − Assess procedures for providing a consistent approach to dose reconstruction, 

regardless of claimants’ exposures by time and employment locations 

Objective 5 − Evaluate procedures with regard to fairness and the extent to which the claimant is 

given the benefit of the doubt when there are unknowns and uncertainties concerning radiation 

exposures 
 

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Objective 6 − Evaluate procedures for their approach to quantifying the uncertainty distribution 

of annual dose estimates that is consistent with and supports a Department of Labor POC 

estimate at the upper 99% confidence level 

Objective 7 − Assess the scientific and technical quality of methods and guidance contained in 

procedures to ensure that they reflect the proper balance between current/consensus scientific 

methods and dose reconstruction efficiency 

6.2.1 Ambiguous Dose Reconstruction Direction 

Direction provided in the site profile as a stand-alone document can be confusing, and in some 

cases, directions are inconsistent throughout a particular TBD. With OTIB-0034 (Kennedy 

2005) for internal co-worker dosimetry data, OTIB-0021 (Merwin 2004) for external co-worker 

dosimetry data, and internal dose estimates for facilities with air sampling programs, the 

approach is somewhat more clearly defined. 

The conditions for the application of environmental dose to ORNL employees are not clearly 

defined in the introduction of the Occupational Environmental Dose TBD, as is usually the case 

with other TBDs. There are recommendations for usage throughout the TBD, but concrete 

direction for which workers receive environmental dose is lacking. 

6.2.2 Inconsistencies and Editorial Errors in the Site Profiles 

Site experts identified some minor errors associated with the NIOSH TBDs. 

• 	 The description of how the tandem accelerator operates in Section 2.2.41 of the Site 

Description is not exactly correct. 

• 	 The dates associated with the accelerators are incorrect. 

• 	 With respect to ORELA (Section 2.2.42 of the Site Description TBD), there is one 

neutron-producing target in the facility. The Tantilum Radiator is actually a target. 

The Site Description TBD identifies radionuclides by facility, and includes information on 

incidents and spills that occurred in the facilities. Some, but not all, of this information is 

considered in the Occupational Environmental Dose TBD (Burns 2004b) and the Occupational 

Internal Dose TBD. In particular, there are radionuclides (e.g., medical isotopes) that have not 

been included in the dose calculations. These radionuclides should be evaluated to determine 

their significance to environmental and internal dose. 

There is discussion in the Occupational Internal Dose TBD (Bollenbacher et al. 2006) of the 

levels of naturally occurring environmental uranium and the ratios of isotopes. However, it is 

not clear how NIOSH is going to handle this with regard to levels seen in urinary and fecal 

excretion and the bioassay monitoring data. Section 5.2.3.3, page 19, states that, “Plots of the 

observed uranium excretion distributions for U-234, U-238, and total uranium are provided 

below.” SC&A, however, failed to find these data in the document. 
 

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The Occupational External Dose TBD states that it does not address skin exposure, yet it 

provides shallow dose correction factors. Furthermore, during the SC&A conference call with 

NIOSH, the dose reconstruction representatives indicated the shallow dose is assigned using 

OTIB-0017 (see Attachment 4). This appears to be providing a methodology to assess skin 

exposure. Clarification on shallow dose calculation is needed in the TBD. 

6.3 UNRESOLVED POLICY OR GENERIC TECHNICAL ISSUES 

A number of issues were identified that are common in the ORNL and other site profiles 

reviewed to date and, in some cases, represent potential generic policy issues that transcend any 

individual site profile. These issues may involve the interpretation of existing standards (e.g., 

oro-nasal breathing), how certain critical worker populations should be profiled for historic 

radiation exposure (e.g., construction workers and early workers), and how exposure itself 

should be analyzed (e.g., treatment of incidents and statistical treatment of dose distributions). 

NIOSH indicated that it may develop separate TIBs in order to address these more generic 

issues. The following represents those issues identified in the ORNL and previous site profile 

reviews that, in SC&A’s view, represent transcendent issues that need to be considered by 

NIOSH as unresolved policy or generic technical issues. 

(1) Direction on the applicability of the TBD and/or TIBs to individual dose reconstructions 

is absent. 

(2) Mobility of work force between different areas of the site should be addressed. Site 

expert testimony that many workers moved from one plant to the next is a complicating 

factor. Establishment of an accurate worker history is crucial in such cases. This will be 

especially difficult for family-member claimants. 

(3) Statistical techniques used in the application of the data to individual workers should be 

further considered and substantiated. 

(4) Dose from the production of less predominant radionuclides from accelerator- and 

reactor-activation activities. 

(5) Dose from impurities and/or daughter products in radioactive material received and 

processed at sites should be assessed as a contributory exposure source. 

(6) The significance of various exposure pathways and the assumptions made that influence 

dose contributions need to be considered (most notably) for solubility, oro-nasal 

breathing, and ingestion. 

(7) Analysis needs to be performed regarding how “frequent or routine incidents” should be 

addressed, given the possibility that such “spike” exposures often may be missed by 

routine monitoring as a function of how often and in what manner it was conducted. 
 

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(8) Availability of monitoring records for “transient or outside workers,” e.g., subcontractors, 

construction workers, and visitors, who may have potential exposure while working on or 

visiting a facility should be ascertained. 

(9) Dose to decontamination and decommissioning (D&D) workers should be assessed. 

Many facilities have large-scale D&D operations, which extend back many years. 

Decontamination and decommissioning operations often require working in unknown 

situations, which may provide unique exposure situations. 

(10) Dose from non-traditional chemical forms of radionuclides, such as high-fired oxides and 

tritides, requires evaluation. 

(11) Dose reconstruction for occupational medical exposures remains incomplete. NIOSH 

needs to reconsider the definition to include all forms of medical radiation exposure to 

ensure its considerations are claimant favorable. 

(12) Dose reconstruction for workers involved in nuclear weapons testing who were employed 

by a site other than the test site. 

(13) Quality Assurance on records provided by the site to the NIOSH/ORAU team is 

necessary to ascertain whether complete information is being provided. 
 

 

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ATTACHMENT 1: NIOSH TECHNICAL DOCUMENTS CONSIDERED 

DURING THE REVIEW PROCESS 

Technical Basis Documents 

ORAUT-TKBS-0012-1, Technical Basis Document for Oak Ridge National Laboratory – 

Introduction Rev. 00, August 11, 2004 (Burns 2004a). 

ORAUT-TKBS-0012-2, Technical Basis Document for Oak Ridge National Laboratory – Site 

Description Rev. 01, August 30, 2006 (Fleming 2006b). 

ORAUT-TKBS-0012-3, Technical Basis Document for Oak Ridge National Laboratory – 

Occupational Medical Dose Rev. 01 PC-1, July 21, 2006 (Fleming 2006a). 

ORAUT-TKBS-0012-4, Technical Basis Document for Oak Ridge National Laboratory – 

Occupational Environmental Dose Rev. 00, May 7, 2004 (Burns 2004b). 

ORAUT-TKBS-0012-5, Technical Basis Document for Oak Ridge National Laboratory – 

Occupational Internal Dose Rev. 00 PC-1, May 30, 2006 (Bollenbacher et al. 2006). 

ORAUT-TKBS-0012-6, Technical Basis Document for Oak Ridge National Laboratory – 

Occupational External Dose Rev. 00, August 11, 2004 (Burns and Mohrbacher 2004). 

Technical Support Documents 

OCAS (Office of Compensation of Analysis and Support), 2003, Tritium Calculations with 

IMBA, OCAS-TIB-002, Rev 00, National Institute for Occupational Safety and Health, Centers 

for Disease Control and Prevention, U.S. Department of Health and Human Services, 

Washington, DC April 22, 2003. 

ORAUT-OTIB-0002, Rev. 01 PC-2, (2004), Technical Information Bulletin, Maximum Internal 

Dose Estimates for Certain DOE Complex Claims, Oak Ridge Associated Universities, Oak 

Ridge, Tennessee, May 7, 2004 (Rollins 2004). 

ORAUT-OTIB-0006, (2005), Technical Information Bulletin: Dose Reconstruction from 

Occupationally Related Diagnostic X-ray Procedures, Rev. 03, Oak Ridge Associated 

Universities, Oak Ridge, Tennessee, December 21, 2005 (Kathren and Shockley 2005). 

ORAUT-OTIB-0017, Rev. 01, (2005), Technical Information Bulletin – Interpretation of 

Dosimetry Data for Assignment of Shallow Dose, Oak Ridge Associated Universities, Oak 

Ridge, Tennessee, October 11, 2005 (Merwin 2005). 

ORAUT-OTIB-0018, Rev. 01, (2005), Technical Information Bulletin – Internal Dose 

Overestimates for Facilities with Air Sampling Programs, Oak Ridge Associated Universities, 

Oak Ridge, Tennessee, August 9, 2005 (Brackett and Bihl 2005). 
 

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ORAUT-OTIB-0021, Rev. 00, (2004), Technical Information Bulletin – External Coworker 

Dosimetry Data for the X-10 Site, Oak Ridge Associated Universities, Oak Ridge Tennessee, 

December 29, 2004 (Merwin 2004). 

ORAUT-OTIB-0034, Rev. 00, (2005), Technical Information Bulletin – Internal Coworker 

Dosimetry Data for X-10, Oak Ridge Associated Universities, Oak Ridge Tennessee, December 

13, 2005 (Kennedy 2005). 

ORAUT-PROC-0060, Occupational Onsite Ambient Dose Reconstruction for DOE Sites, Oak 

Ridge Associated Universities, Oak Ridge, Tennessee, June 28, 2006 (Winslow 2006a). 

ORAUT-PROC-0061, Occupational X-ray Dose Reconstruction for DOE Sites, Oak Ridge 

Associated Universities, Oak Ridge, Tennessee, June 21, 2006 (Winslow 2006b). 
 

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ATTACHMENT 2: SITE EXPERT INTERVIEW SUMMARY 

<RESERVED> 
 

 

 

 

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ATTACHMENT 3: KEY QUESTIONS AND RESPONSES FOR 

NIOSH/ORAU REGARDING SITE PROFILE DOCUMENTS 

SC&A submitted questions related to the ORNL Site Profile documents to NIOSH on August 1, 

2006. The NIOSH/ORAU team provided written responses to these questions on August 25, 

2006. Both the questions and responses have been provided below. A summary of the 

conference call conducted between SC&A and the NIOSH/ORAU team with regard to these 

questions is presented in Attachment 4. 

Questions and Responses 

Site Description (ORAUT-TKBS-0012-2) 

(1) How will potential exposures to site personnel from unfiltered emissions be handled for 

claimants? Page 21 of the Site Description TBD, ORAUT-TKBS-0012-2, says that 

airborne exhausts from B 3001/3018 did not include a filter system before 1948. Later 

they found particulates still airborne from fuel slugs pushed out back of the reactor. 

Also, gaseous wastes, such as noble gases Ar-41, Xenon/Krypton, and radioiodines, were 

not filtered or quantified. 

These issues are addressed in the onsite environmental TBD. 

Airborne wastes were removed from Building 3001 and exhausted to the 

atmosphere from a 200-ft stack (Building 3018) immediately north of the 

Graphite Reactor Building..... 

(2) Have interviews been conducted with workers at the Y-12 ORNL Biological Sciences 

division to determine if any problems with sources occurred? (See pages 29 and 30 of 

Site Description ORAUT-TKBS-0012-2). 

Information of this nature is gleaned through the Computerized Assisted 

Telephone Interviews (CATIs), which seeks input from claimants and Worker 

Outreach process, which seeks input from all workers and former workers, 

claimants and potential claimants. The TBD team did not seek out workers 

individually. 
 

 

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(3) How is dose reconstruction going to be possible for numerous buildings where the dates 

of operations and processes are not even known? (See page 26 of the Site Description, 

ORAUT-TKBS-0012-2). This could be an issue when the site relied on the area health 

physicists to determine if in-vitro bioassay monitoring and in-vivo monitoring should be 

done, and appears only to be done if there were known spill/incident, or air and 

contamination sample results that indicated it should be done, yet the dates of different 

operations in the areas is not known. 

Available information including individual information regarding individual 

bioassay and information from the CATIs is being used to reconstruct doses. 

When a dose reconstruction warrants, coworker bioassay will also be considered. 

Because the bioassay data were collected in response to incidents or at times 

when results would likely be the larger, their use in coworker models and chronic 

exposure scenarios is considered claimant favorable. (The site description 

section is not meant to be the guide for how dose reconstructions will be done, 

rather it is provided to describe the sites chronological activities and source 

terms that are important to understanding the opportunities for employees to 

incur radiation exposure. 

Occupational Medical Dose (ORAUT-TKBS-0012-3) 

(1) The Occupational Medical Dose TBD, ORAUT-TKBS-0012-3 (Fleming 2006a) states in 

the introduction, in Section 3.1, that all internal and external dosimetry monitoring results 

are considered as part of a valid dose reconstruction. However, it excludes radiation from 

radon and its daughters, and any radiation from diagnostic x-rays received to treat workrelated injuries. Explain the procedures that NIOSH plans to adopt to subtract such 

radiation dose when the dosimeter was inadvertently exposed during x-ray procedures, or 

due to naturally occurring radiation exposure in the workplace? 

This passage in the TBD has not been accurately summarized. Note that exposure 

from naturally occurring radon in conventional structures [and the environment] 

are not considered “occupationally derived.” (Radon is an issue for the 

environmental or internal dose section.) Unless information is available that 

specifically indicates that a dosimeter’s reading was due to non-occupationally 

derived exposures, no attempt would be made to adjust the reported result for this 

possibility. Such instances would be handled on a case-by-case basis. To date, 

we are unaware of dosimeters being identified as worn during medical 

examinations and used to monitor occupational work radiation doses. 

(2) The TBD (Fleming 2006a) in Section 3.2 states that preplacement, annual, and 

termination examinations are all contributors to the occupational medical dose, as defined 

under EEOICPA. Would NIOSH define the nature and type of diagnostic x-rays that 

constitute “annual” examinations? Would NIOSH also explain and clarify whether 

periodic screening exams, such as for TB or qualification screenings for respirator 

certification and asbestos workers, constitute annual exams? 
 

 

 

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Occupationally required (versus medically required) x-rays are included in dose 

reconstruction. Periodic occupationally required screening X-ray exams that are 

performed because of statute or policy, and that are not based on an employee’s 

individual need for diagnosis are generally considered to be in this category, and 

the SC&A examples of screening exams for TB, respirator certification, and 

asbestos work fit this category. 

(3) The TBD (Fleming 2006a) in Section 3.3 indicates that exams required as a “condition of 

employment” only are covered. Could NIOSH further define and clarify what constitutes 

a “condition of employment?” Most x-ray procedures are given on a voluntary basis, 

such as routine annual chest x-rays. Does this constitute a “condition of employment?” 

The phrase "condition of employment" comes from 42 CFR 81, where there is a 

little more context to the phrase. 42 CFR 81 states that "Additionally, some 

employees covered by EEOICPA were required, as a condition of employment, to 

undergo routine medical screening with x-rays. The dose resulting from these 

x-rays will be included in their dose reconstruction." 

The ORAUT has used an OSHA definition of medical screening as tests performed 

on asymptomatic persons, without history or complaint. We try to research the 

site's documents to uncover screening protocols, study claim files to determine 

exam frequencies, and search open literature for common practices for 

comparable time periods in order to determine which x-ray procedures and 

projections should be considered screening for a particular DOE site. 

(4) The TBD states in Section 3.3 that photofluorographic (PFG) exams were used only for 

preplacement screening from 1943 through October 3, 1947. The TBD infers that only 

one PFG exam was given to each new employee. However, available historical 

documents indicate that multiple uses of PFGs over several years were given to some 

employees. If the doctor requested further exams, a follow-up regular chest x-ray was 

also given. Would NIOSH clarify how the dose reconstructor ought to consider 

cumulative dose from PFG exams, particularly for claimants who worked during 1943– 

1947? 

Page 11 of the current TBD indicates that a stereoscopic PFG for pre-placement 

should be assumed for individuals who began employment at ORNL from 1943 

through 1947, and that documented retakes, routine annuals, and terminations 

that may have occurred during that time period would have been a 14-in x 17-in 

radiographic PA images provided at the Oak Ridge Hospital. 

A sample of about 15% of the entire set of PFGs located at ORNL were reviewed, 

and in only one instance was an individual found to have received more than one 

PFG, and these were taken one year apart. While additional PFG sets may exist 

for an individual, it appears that this was not the norm. 
 

 

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(5) The TBD in Section 3.4.1 discusses the peak applied voltage and filtration used in the 

Picker Model R-2 installed in October 1947. The TBD indicates that the medical staff 

asserted that the inherent filtration was 0.04 mm of aluminum (Al). Further review of 

historical documents suggests that the 0.04 mm Al value is actually derived from surveys 

of the unit by R.L. Tuck in 1956 and 1957, and appears to be in error. Later 

measurements, taken by E. Gupton in 1957, provide an assumed value of 1.0 mm Al, 

with no basis being given. Could NIOSH clarify the actual filtration for this unit? The 

importance to be associated with this question is that the entrance skin exposures from a 

PA chest of 21 mrad may be incorrect if the assumed filtration and half-value layer 

(HVL) is in error. 

Though it is acknowledged that the filtration provided by Mr. T. L. Tuck both in 

2003 and documented on several surveys conducted in the mid-1950s for the 

Picker Model R-2 unit used at ORNL was 0.04 mm of Al, this filtration value does 

not appear to be reliable. On those same “historical documents” referenced 

above, the measured skin dose was 25 mrad at ORNL. Also in the two peerreviewed reports written by T. A. Lincoln and E. D. Gupton that “assume” a 

1.0 mm filtration, the measured exposure to the skin at ORNL for a PA CXR was 

21 mrad. Both these measured values are less than the entrance skin exposure 

(56 mrad) used in the TBD to estimate organ doses and as indicated in the 

footnotes to Table 3-4, “This value is approximately a factor of 2.7 greater than 

that measured and is claimant favorable.” Based upon a comparison of the 

measured vs. calculated ESE values, using a filtration of 1.0 mm Al is favorable 

to the claimant. 

(6) The TBD in Section 3.4.2 states that PFGs were not performed at the ORNL medical 

facility. What we need to know is if any employees were sent to the Oak Ridge Hospital 

for chest radiography either as preplacement or annual exams after October 1947? Has 

NIOSH identified when the PFG unit was no longer used at the Oak Ridge Hospital? 

Other documents referenced in the TBD (Lincoln and Gupton 1958a) and (Lincoln and 

Gupton 1958b) suggest that employee x-rays may have occurred at two hospitals, two 

local radiology offices, and other Oak Ridge sites. Can NIOSH establish whether other 

facilities may have had PFG units other than Oak Ridge Hospital? 

The reviewed information did not indicate ORNL employees received 

preplacement or annual chest exams at Oak Ridge hospital after October 1947. 

The date for discontinuance of the PFG unit at the hospital and whether PFG 

units were used at other locations is currently irrelevant to the ORNL TBD and 

dose reconstructions, given these facilities were not used for the occupationally 

required medical x-rays. The Lincoln and Gupton articles do not indicate that 

ORNL’s occupationally required medical x-ray examinations were made at 2 

hospitals, 2 radiology offices and other Oak Ridge sites, rather they state that 

these researchers wanted to account for all x-ray doses received by the ORNL 

population. A review of the articles confirms this as doses from exam-types that 

are not associated with occupational requirements, such as abdomen and knee 

exams, are tabulated in the articles. 
 

 

 

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(7) The TBD in Section 3.4.4 states, according to a long-time radiology technician, that x-ray 

beams at ORNL were always well-collimated. However, later in Section 3.5, the TBD 

states that lead-lined gonad shields were used at ORNL until they were discontinued in 

favor of collimation of the beam. Could NIOSH clarify this apparent discrepancy in the 

TBD? 

Reviewer is directed to the discussion on page 10 that describes this issue. 

Though NIOSH had verbal (from the radiological technician) and documented 

(20 cm cone shield from the Lincoln and Gupton papers) information that the 

medical radiographic equipment was well collimated, poor collimation was 

assumed until 1963. This decision by the ORAUT was based on an analysis of 

organ doses in a table in the Lincoln/Gupton papers. The TBD states the results 

of this analysis “indicate that the dose conversion factors (DCFs) given in 

ICRP 34 for a well collimated beam may not apply and therefore uncollimated 

DCFs found in ORAUT-OTIB-0006 ... were used until a newer x-ray unit was 

procured in 1963. Table 3-3 also indicates that in the assumed operating 

parameters to “Use uncollimated DCFs for organ dose estimates for the Oak 

Ridge Hospital and Picker units. Collimation of the beam for the Westinghouse 

and Bennett units is assumed to be good and ICRP 34 values used.” Though 

information obtained from the site indicates that the beam was collimated, the 

decision of using DCFs from uncollimated beams has some bases and is favorable 

to the claimant. 

(8) The TBD in Section 3.5 states that dose reconstructors need only consider chest PA and 

LAT x-rays, and the lumbar spine series (4 films) in a claimant dose reconstruction. As 

also evidenced in Table 3-2, it appears that all dose calculations prior to 1996 are to be 

based on a PA chest view only and not a LAT. Can NIOSH clarify and document that 

annual chest exams did not include a LAT view, unless specifically prescribed by the 

physician? 

The information that was provided by ORNL and limited x-ray films that were 

reviewed indicated that chest x-rays were conducted until April 18, 1996 with one 

PA projection. Table 3-2 indicates date ranges, techniques utilized, and what 

people were involved in the medical x-ray program at ORNL. The table identifies 

PFG chest x-rays, conventional PA chest x-rays, lumbar spine series, and lateral 

chest x-rays that were given at ORNL. This information was provided to NIOSH 

from ORNL. 

(9) The TBD in Table 3-2 lists the type of equipment and frequency of exams. The table 

indicates that chest radiography was only provided as a preplacement requirement after 

1963, and annual exams were not required. Can NIOSH clarify this table, relative to 

statements in the TBD that annual chest x-rays were only required until 1970, after which 

they could be waived? Can NIOSH document that respirator program employees only 

required an exam every 3 years after 1976? 
 

 

 

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The text indicates that the annual and termination X-ray examinations were 

provided into the 1970s; although there are inconsistencies among the text, 

Table 3-2 and Table 3-4, dose reconstructors are assuming preplacement, annual 

and termination examinations through 1976. The reference regarding the X-ray 

examination requirement for respirators workers, Ref ID 10114, can be found in 

the SRDB. 

(10) According to other historical documents, it appears that extra doses may have resulted 

from lining up several employees inside the x-ray room to await annual x-rays. Has 

NIOSH attempted to estimate these potential extra doses to workers who were exposed 

by this practice? 

If such a practice did occur, the exposure from scatter of the beam, which would 

have been much smaller (probably on the order of 1% or less), would be 

adequately accounted for in the uncertainty assignments associated with 

reconstructing x-ray doses. 

Occupational Environmental Dose (ORAUT-TKBS-0012-4) 

(1) Could NIOSH refer to the instructions or guidelines used to reconstruct missed 

environmental doses? We recognize that Attachments 4A, 4B, and 4C provide annual 

airborne release data, annual average airborne concentration and intake data, and annual 

average site-wide exposure rates data. 

Environmental doses are only used to estimate doses for unmonitored workers, 

who would not have required radiological monitoring. We are not sure exactly 

what the question is; we are not aware of what “missed” environmental doses are 

being questioned. The only missed dose mentioned in this TBD is for external 

dose and this doesn’t appear to be the issue. Please clarify question. 

(2) Could NIOSH clarify how they plan to identify and deal with those potential 

unmonitored workers that may receive missed environmental doses? Are they identified 

by their working location or job title? Are there monitored workers who would leave 

their TLD in their working location before they left ORNL site? Were the monitored 

workers required to receive whole-body counting and bioassay testing routinely (weekly 

monthly, and quarterly)? If not, how do we know whether these monitored workers 

received missed environmental doses? 

OTIB-0014 gives guidance for determining when to assign environmental intakes 

rather than internal dose. 

Environmental doses are assigned for unmonitored exposures. If bioassays or 

other monitoring records provide indications for a given type of radiological 

exposure, the assignment of dose based on environmental estimates is not 

appropriate. Therefore the questions regarding bioassays are considered 

irrelevant in the context of this question. 
 

 

 

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It is unclear what “missed” environmental dose means. Workers who have no 

monitoring may or may not be identified by location or job title. Where 

monitored workers left their dosimeters is not relevant to the Occupational 

Environmental Dose TBD. Bioassay information for monitored workers is 

addressed in ORAUT-TKBS-0012-5). 

Regarding environmental dose assignments for unmonitored employees, the 

maximum of the site wide annual median environmental intakes, which includes a 

favorable to the claimant assumption of 2000 hours of exposure at the maximum 

of the annual median air concentrations is used to estimate unmonitored worker 

internal doses. This maximizing approach is sufficiently large to address any 

“missed” internal dose. 

(3) We suggest that in future updates to the Occupational Environmental Dose TBD (Burns 

2004), a section that addresses accidental spill of radioactive material or waste and how 

the dose reconstructor would determine a dose to involved workers? Could there not be a 

source of missed environmental dose to workers by way of contaminated soil - 

windblown of resuspended contaminated soil? 

The environmental section of the TBD is used to assign dose to workers who 

would not have been subject to site and regulatory monitoring requirements, so 

these workers’ involvement in spills would not be routine. Any airborne 

contaminated soil would have been indicated by the onsite air sampling network. 

Information on spills and the like is included in the site facilities TBD. 

(4) Could NIOSH comment on potential dose that might be expected from worker exposure 

when working around evaporation ponds or cooling ponds? The Occupational 

Environmental Dose TBD (Burns 2004b) mentions the presence of the settling basin that 

contributed to the missed environmental dose to workers, but there is no mention of 

evaporation ponds or cooling ponds. 

It’s not clear what evaporation/cooling ponds this question refers to. 

(5) Has NIOSH considered dose reconstruction for exposure received during uranium ore 

processing, ore contamination, tailing piles, or airborne dust concentration data from the 

ore processing? 

Although we recently learned that the uranium ore and residues and scraps from 

uranium operations were stored in at least two of the warehouses at the Clinton 

Lab, it is not clear that any uranium processing was taking place (there was 

mention of a rebagging process) or that uranium ore tailings piles were onsite. 

Further investigation of this source term will be pursued for reconstructing dose. 

(6) Has NIOSH considered the impact of waste piles and waste drums as potential sources of 

exposure in the Occupational Environmental Dose TBD (Burns 2004b)? 
 

 

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If there were such collections of waste which contributed to the ambient exposure 

rate at X-10, currently they are considered addressed via the monitoring data 

used in the TBD. 

(7) Did NIOSH identify other important sources of environmental exposure, other than 

airborne releases from stacks and reactors referenced in Burns 2004b? Also, would 

NIOSH explain how they plan to provide guidance to the dose reconstructor for these 

additional sources of environmental exposure? 

The current TBD addressed what we believed to be the major contributors to 

exposure, but as additional information becomes available and if it indicates 

modifications are warranted, information will be added to the TBD for use by 

dose reconstructors. 

(8) Could NIOSH provide clarifications on the reasons for not including any discussion or 

any detailed uncertainty analysis of airborne concentration data and exposure data? Have 

these uncertainties been quantified and are they utilized in doing environmental dose 

reconstructions? 

Uncertainties in the form of geometric standard deviations are asserted in the 

attachments summarizing the exposure data. A default GSD = 3 is assigned to 

environmental doses. 

(9) The Occupational Environmental Dose TBD (Burns 2004b) indicated that the annual 

"average" airborne concentrations and annual "average" exposure rates are “bounding” 

values. The TBD, however, indicated that these site-wide average data might not be 

appropriate for cases in which a composite exposure profile for an unmonitored 

individual is inconsistent with a known or asserted exposure history. Would NIOSH then 

explain why these site-wide average data can be considered to be “bounding?” Is there a 

more claimant-favorable way to deal with this? 

This is explained in the TBD. The method used is believed to be extremely 

favorable to claimants since it includes contributions from sources in areas which 

should have been off-limits to unmonitored persons. Monitoring data from more 

remote locations (which in reality are likely more indicative of the exposure to an 

unmonitored individual) were not included in the averages, resulting in a high 

(favorable) bias. 

(10) Some location-specific data were not provided in the TBD. For example, on page 33 of 

the Occupational Environmental Dose TBD (Burns 2004b), specific site locations (e.g., 

706 D) were not provided for use by the dose reconstructor. Would NIOSH provide the 

location of these data? And clarify how can a dose reconstructor find these data to use in 

the dose reconstruction efforts? 

The data appear in the routine HP Division progress reports and similar 

documents and dose reconstructors have access to TBD authors if they need 
 

 

 

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assistance in locating site data. Keep in mind these locations were inside the 

protected area and thus should not have been accessible to unmonitored workers. 

However, specific locations of workers are not known, so the citation of specific 

areas does not generally provide the DR with information that can be applied to a 

particular case. 

(11) The Occupational Environmental Dose TBD (Burns 2004b) focuses on onsite airborne 

I-131 concentration, onsite airborne concentration of mixed fission products (MFPs), 

onsite airborne concentrations of tritium, and onsite exposure rate data. Are there any 

MAPs from the different reactors? 

These are included in the MFP data, which are gross beta. MFP should probably 

be changed to mixed fission/activation products (MFAP). 

(12) Why has NIOSH chosen the breathing rate of 1.2 m3/hr? Is that value claimant favorable 

for maintenance workers and movers? 

The chosen breathing rate is standard for routine analyses, and when combined 

with other assumptions in the TBD (2000-hour exposures at the maximum of the 

median annual intake rates), we believe the resulting intakes estimated for 

unmonitored workers are claimant favorable. The issue of breathing rate is a 

programmatic issue and not a site profile issue. 

Occupational Internal Dose (ORAUT-TKBS-0012-5) 

(1) How would NIOSH plan to handle dose reconstruction for the large numbers of 

unmonitored workers up through the 1970s, 1980s or maybe even 1990s?
 

On page 11 of the ORNL Internal Dose TBD (Bollenbacher et al. 2006), the TBD notes 

that “urine samples were collected in the early years of the bioassay program based on 

the area health physicist’s knowledge of field conditions (e.g., known spills/incidents, air 

and contamination sample results, etc.). This practice of scheduling did not utilize a 

specified sampling frequency (Auxier 2004; Henley 2004).” It is further noted on page 21 

that “Although several located documents stated that baseline and specified monitoring 

frequencies were utilized to make in-vivo measurements, Berger (2003) and McLaughlin 

(2004) indicated that a full in-vivo monitoring program did not exist at ORNL until 

approximately 1994, when site internal dosimetrists became responsible for identifying 

personnel for counting.” This practice for determining sampling for both in-vitro and invivo monitoring throughout the history of the site up through the 1980s or even 1990s 

raises a serious question of unmonitored workers with potential for uptakes. The issue is 

further exacerbated in Table 5-3 (page 11), which states that these sampling frequencies 

should be followed, but the text notes that these may not have been followed. Table 5-3, 

last row, also notes for “All others – Consult w/Internal Dose Group.” What was used 

for the hundreds of other radioisotopes? 
 

 

 

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This issue is present at almost all sites. For long-lived nuclides, routine (i.e., 

samples collected frequently on a specific schedule) are not necessary for 

performing a bounding dose reconstruction; the material is retained long enough 

that a few samples can provide enough information. For gaps after the last 

bioassay sample or for shorter lived nuclides, coworker data would be used. 

(2) Clarify NIOSH’s approach in determining tritium exposure. Tables 5-1 on page 10 notes 

over 2,000 samples were analyzed. Were there tritides involved or handled at ORNL? 

There is no discussion of this potential for exposure or guidance to the dose 

reconstructor. 

At this point it is not clear if metal tritides were handled at ORNL, and if they 

were what quantities and what processes were involved. In general tritium tritide 

contributes little dose unless there are large quantities of contamination found in 

the work place. Currently tritium dose is calculated and assigned assuming 

exposure to HTO whenever monitoring for it is found in the dosimetry record. 

Additional information regarding ORNL tritium use will be sought. 

(3) How is NIOSH going to deal with the conflicting assumptions for solubility classification 

for uranium? The ORNL Internal Dose TBD, ORAUT-TKBS-0012-5, states that the 

default should be Type S (page 67), however, the LANL Internal Dose TBD, ORAUTTKBS-0010-5, recommends using Type M. The ORNL Internal Dose TBD does not 

address the Integrated Modules for Bioassay Analysis (IMBA) NIOSH Phase I database 

USDOE Version 1.0.42, Tables 5-11 and 5-12 for isotopic composition for uranium. 

Why not? 

No citation on page 67, stating the uranium default absorption type is type S, was 

located. Dose reconstructors are assuming ORNL exposures were to type F, M, 

or S uranium, and the selection is based on claimant favorability (Section 5.1.2 of 

the TBD). The TBD provides site specific information and not information on 

dose reconstruction tools, like IMBA (not a database), which are addressed in 

other documents or guidance; however, this question might be referring to 

enrichment assumptions for uranium at ORNL. For uranium bioassays, where 

measurements were based on gross alpha counting, knowledge regarding the 

enrichment is not needed to reconstruct dose. 

(4) The ORAUT-TKBS-0012-5 TBD, Section 5.2.3.3, page 18, talks about the 

environmental uranium, but I cannot find any reference to the fact that it was subtracted 

from the bioassay results. Were the background uranium values subtracted from the 

bioassay results? 

These data didn’t exist until the 1990s and are apparently used by ORNL for an 

indication to investigate suspect uranium results. It does not appear they were 

used for the purpose of background subtraction. There is no indication that 

estimates of environmental uranium in urine were subtracted from bioassay 
 

 

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results reported by ORNL and such an adjustment is not being made in this dose 

reconstruction project. 

(5) The ORNL TBD (Bollenbacher et al. 2006) discusses background environmental 

uranium and fallout Cs-137, but there is no mention of a problem or issue of radon or 

radon daughters. How were these handled by the dosimetrists? What kinds of exposures 

to these occurred and are they accounted for by the NIOSH dose reconstructors? 

Radon and radon progeny interference is typically not an issue for internal 

dosimetry measurements (although it might be a consideration during design and 

set up of counting installations, as well as determination of parameters related to 

determining detection thresholds). Occupation radon exposure information for 

ORNL will be gathered for incorporation into the site profile; see response to 

Question 6 of this section. 

(6) Early documents (Memoranda to Col. Stafford L. Warren, March–June 1944, by John 

Ferry, et al.) certainly indicate that radon and radon daughters were an issue for the 

ORNL site. How are the unmonitored exposures to these radionuclides going to be dealt 

with by dose reconstructors? There is no reference or discussion of this problem as a 

potential internal exposure issue in the ORNL Internal Dose TBD. This has been an issue 

of concern at most all other DOE sites. 

From available references, it appears that measured radon levels in 1944 ranged 

from 240 to 7276 pCi/L in storage warehouses (2 and 3) that contained uranium 

ore at the Clinton Engineer Works 0101 area. We agree that radon exposures 

need to be addressed in the ORNL TBD and certain dose reconstructions. 

(7) Information provided on page 9 (Bollenbacher et al. 2006), Section 5.1.1, and on page 12, 

Section 5.1.5, is somewhat confusing. It is stated on page 9 that, “Table 5-1 lists 

radionuclides included in in-vitro bioassay results provided by ORNL for the period from 

1947 to 1988,” and further points out on page 9 that “These values do not reflect the total 

number of in-vitro bioassays performed by ORNL in this period, because not all hardcopy records are in the database.” Page 11 states that, “...ORNL has entered much 

historical in vitro monitoring data into a database.” Table 5-1 has numerous 

radionuclides listed with sample sizes of 1 to 10. With all the hundreds of radioisotopes 

handled at ORNL, as noted on page 30 of the Site Description TBD, ORAUT-TKBS0012-02, how many in-vitro bioassay samples for these hundreds of radioisotopes really 

exist? (Emphasis added.) 

The reference to page 30 of ORAUT-TKBS-0012-2, which includes mention of 

250 isotopes, is in a paragraph that includes description of the Stable Isotopes 

Program, and the number 250 does not refer only to radioactive isotopes, but 

includes both radioactive and stable isotopes. We have a database of all of the 

results, if this information is needed. However, this question does not appear 

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(8) What guidance is given to dose reconstructors regarding Am-241 exposures? The TBD 

Section 5.2.3.1, page 18 (Bollenbacher et al. 2006), notes that, “...monitoring of 

transplutonium elements was unable to differentiate between such nuclides as 241Am and 

244 Cm. The default radionuclide to use with measurements involving trivalent alpha 

actinides would be 241Am. The detection sensitivity of the transplutonium analysis 

technique is not well documented for samples processed before 1985.” This seems like a 

weak approach considering Cm-244 was handled as a radionuclide. There is likewise no 

discussion of the ingrowth of Am-241 and the issues it presents in using in-vitro and invivo monitoring for internal dose reconstruction. Table 5-10 (page 17) includes Am-241 

in the sequential analysis; however, americium should not be inferred as being absent 

without the presence of plutonium. Also, when plutonium mixtures are present, the 

absence of americium above detection levels should not preclude the calculation of dose 

from americium contribution to the mixture. The TBD does not appear to address the 

issue of missed dose from americium. 

Because the TBD gives no guidance on plutonium mixtures, the DRs don’t include 

Am in the Pu intakes; it would only be included if there were Am results. 

According to Dave McLaughlin (ORNL internal dosimetrist) who said they make 

no such assumptions either – they treat Pu and Am independently because they 

have Am sources unrelated to Pu. However, it would be helpful to know whether 

weapons-grade, fuel-grade, or no specific Pu mixtures are being used. 

(9) Bioassay sampling in early years, before routine sampling was formalized, depended on 

area health physicists’ decisions based on spills, incidents, and air and contamination 

sample results. Many of the radioisotopes in Table 5-1 (page 10) have less than 10 

samples recorded, and this is only a small fraction of the more than 250 radioisotopes 

noted on page 30 of the Site Description TBD, ORAUT-TKBS-0012-2. What are dose 

reconstructors to do for these poorly recorded radionuclides? Or, can it be proven that no 

potential for exposures occurred for the radionuclides? Certainly there were indications 

that some contaminants got to the environment, as noted in the Site Description, and it is 

hard to imagine that none of the workers received exposures. 

As noted in response to Question 7 of this section, the Site Description TBD 

referred to 250 isotopes, not 250 radioisotopes. Not all of these isotopes are 

radioactive. It’s not clear what a poorly recorded radionuclide is. Given that 

there are bioassays, and that monitoring may be indicative of a potential for 

exposure, a conclusion of no exposure potential cannot be drawn for all workers, 

although for many workers this might be the case. For the early years, the TBD 

page change that provides instruction to use ORAUT-OTIB-0018 addresses 

exposure to mixed source terms. 

(10) What are the boundary conditions for chronic intakes of radionuclides? The Internal 

Dose TBD does not address this issue or the maximum allowable concentrations (MAC) 

for potential missed dose calculations. A method for identifying workers and assigning 

missed dose for those potentially exposed to all the assorted radionuclides for which 

MDAs have been determined is lacking in this document. Likewise, for all the other 
 

 

 

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radionuclides, such as those produced by the ORNL Isotopes Production Group, a 

method is also lacking for assigning a missed dose. 

It would not be reasonable to assign missed dose (estimated from the detection 

threshold of the monitoring result for an individual) from all the radionuclides 

listed to all potentially exposed workers who did not have associated bioassay. 

For many of the listed radionuclides, the amounts and periods of use were limited 

and the associated dose conversion factors. Where there is evidence that 

monitoring should have occurred for a particular radionuclide and a particular 

claim, the dose from the particular radionuclide is reconstructed by the dose 

reconstructor, or the method of estimating internal dose is sufficiently favorable 

to the claimant that the dose is adequately accounted for, or the additional dose is 

ignored because it will not influence the compensability decision. 

(11) The ORNL Internal Dose TBD (Bollenbacher et al. 2006) does not address or consider 

potential exposure from the ingestion pathway, especially in the early years. Many of the 

radionuclides handled at ORNL were highly soluble. What guidance will be given to 

dose reconstructor for calculating this missed dose? 

Bioassay is an indicator of all modes of intake and the assumption of inhalation 

intakes (when no specific information is available, and which include a portion of 

ingested material in the inhalation model) is generally a claimant-favorable 

assumption, and adequately addresses dose, especially when distributions are 

assigned to account for uncertainty in the modeling. For doses derived from air 

concentrations in the workplace an ingestion dose would be estimated in 

accordance with NIOSH guidance. 

Missed dose (the dose that is missed because detection capabilities related to the 

specific item being monitored) is assigned by dose reconstructors according to 

generic project guidance. We believe this question might refer to unmonitored 

individual dose. For the early years, intakes are estimated using the ORAUTOTIB-0018 approach as identified in the ORAUT-TKBS-0012-5 Rev 00 PC-1. A 

coworker model is also available for estimating dose in ORAUT-OTIB-0034. 

Because both of these models are believed to be sufficiently generous in assigning 

dose to their intended population, workers who were unmonitored, ingestion is 

not specifically included. 

(12) Will some of the processes involved at ORNL be treated as high-fired Super S forms of 

plutonium? It is stated on page 17 and 22 of the Site Description TBD, ORAUT-TKBS0012-2, “The Fuel Cycle Alpha Facility (FCAF) fabricated PuO and (Pu,U)O fuel pellets 

and was used for coating sol gel-derived microspheres with pyrolytic carbon.” This 

occurred in B 3019 and 4508. It is also noted that gloveboxes were equipped to 

synthesize, press, and sinter pellets of Pu and U nitrides in B 4508. 

Assumption of absorption types is considered a generic dose reconstruction issue. 

The site profiles identify the radioactive materials and associated processes, so 
 

 

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appropriate absorption types can be determined and applied by dose 

reconstuctors. It appears that when guidance for assigning super S plutonium 

intakes is finalized, at least some ORNL dose reconstructions will be impacted. 

Occupational External Dose (ORAUT-TKBS-0012-6) 

Monitoring 

(1) Page 11 (Burns and Mohrbacher 2004) – It is stated that workers required to work in 

restricted areas more than 3 days per week were assigned beta-gamma monitors. Do the 

workers who only worked in restricted area 1 or 2 days per week have any dose of 

record? Could a worker work in numerous different restricted areas 1 or 2 times each 

week, and not be monitored because he/she did not work in any given restricted area 

more than 1 or 2 days in a week? 

Early exposure limits were based on daily (0.1 R/day) or weekly (0.3 R/wk) limits 

and typically monitoring of workers expected to exceed perhaps 10% of these 

limits. There is little distinction between assigning pocket dosimeters to workers 

routinely entering restricted areas and workers using so-called “trip dosimeters”, 

i.e., those issued on an ad hoc basis, whenever there is an occasional entry into a 

restricted area. 

(2) Page 11 (Burns and Mohrbacher 2004) – The last paragraph states that as of 1951, all 

regular workers were required to wear a combination security badge and film dosimeter. 

Did this include NTA film in all cases? 

The design of that badge had the NTA film behind the beta/gamma film, and often 

a positive result greater than a defined level on the more easily processed 

beta/photon film was a pre-requisite to processing the NTA, so yes, the NTA was 

normally present. It wasn’t always processed, however. Neutron dose is 

assigned using the guidance in the technical basis document, using recorded 

results and neutron:photon ratios. Due to dosimetry limitations, both are 

discussed in the TBD and used in dose reconstruction. 

(3) Page 17 (Burns and Mohrbacher 2004) – It is stated that zero and blanks do not mean the 

same thing on the individual data cards as they do on the computer system(s), and it is 

recommended that the dose reconstructor use the individual data cards if available. Are 

these data cards frequently available? Please discuss how zero and blank entries are 

handled during DR, especially as dosimetry/recording systems changed throughout the 

years. 

A review of the NOCTS files indicates the cards are routinely available, even in 

the years when the computer printouts are also included. Dose reconstructors 

assign a missed or estimated dose using OCAS-IG-001 guidance whenever there 

is not a positive recorded dose for a routine exchange period, and it is believed 

likely that the individual was incurring exposure. 
 

 

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(4) Pages 60 and 61 (Burns and Mohrbacher 2004) – The maximum readable dose on gamma 

films was set at 5 R to 20 R. How were doses assigned for workers whose films were 

blackened? Did this occur very often? 

Some records contained evidence of blackened films. The technician selected and 

recorded various codes when this happened (valid exposure, light leak, water 

intrusion, etc.). The ORNL HPs then performed an investigation and assigned a 

dose. This documentation is included in the workers' records. 

Non-penetrating radiation doses 

(1) Page 12 (Burns and Mohrbacher 2004) – It is stated that ORNL did not report skin (or 

superficial) dose until the second half of 1961. How will skin, breast, and testes DR be 

performed for 1943–1961? 

The open window (OW) data should be used in assigning dose. Those dose 

quantities were not used prior to 1961, but that does not mean that data were not 

available to make an evaluation. 

(2) Page 19 (Burns and Mohrbacher 2004) – Beta dosimetry is mentioned in Section 6.3.3.1, 

which discusses some shortcomings/unknowns in the early years. Has any effort been 

made to identify and list the major job titles and/or locations where beta exposures should 

be of particular concern during DR? 

No, dose reconstructors routinely use information provided in the respective 

claims, in the claimant interviews, and in the DOE provided dose records to 

evaluate significant parameters potentially impacting dose reconstruction. As 

noted in the TBD, a quantitative assessment of workplace beta and photon 

radiation fields has not been located. As such, the dose reconstructor has little 

option but to assume favorable to the claimant values. 

Neutron doses 

(1) Page 16 (Burns and Mohrbacher 2004) – Table 6-2 lists NTA film in routine use around 

1949. How will neutron DR be performed for 1943–1949, if neutron dose information 

does not exist to determine n-p values for that period? 

This issue was recognized when the TBD was developed and is one reason to 

select the use of a neutron-to-photon (n:γ) ratio approach to assign neutron dose. 

The intent of the TBD guidance was to provide n:γ dose information for ORNL 

reactors so dose reconstructors could apply these ratios to the early time period 

(2) Page 21 (Burns and Mohrbacher 2004) – It is stated that it is unclear how thermal and 

fast neutrons were reported using NTA film with open window (OW) and cadmium (Cd) 

filters. Has any information been found that indicates that thermal neutrons were 
 

 

 

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recorded in the dose records? If so, how were they measured, and what method was used 

for calibration? 

Limitations of NTA to measure thermal neutrons were well known, but portable 

radiation instrumentation could be readily used, and certainly activation of 

cadmium with subsequent photon emission was a common dosimetry practice. 

The graphite reactor (experimental ports with appropriate filters) may have been 

used for calibration. Since it was a research tool, they had well-characterized 

beams. Thermal neutron dose may not have been reported on the Kardex cards 

(which may be what prompted the statement in the TBD to begin with). There are 

measurements to differentiate fast and thermal neutron components of graphite 

reactor beams and determinations of how the film badge responded to these. 

They had a tolerance limit for thermal neutrons even though they couldn’t 

measure them. 

(3) Page 24 (Burns and Mohrbacher 2004) – It states in the last paragraph that the conditions 

concerning neutron exposures are likely unchanged over time. This is a broad statement, 

especially in a research environment. What investigation and documentation supports 

this concept? 

The TBD states “Conditions at reactor, accelerator, and calibration facilities are 

not likely to have changed significantly over time, and operations at the 

Radiochemical Engineering Development Center (REDC) have remained 

consistent over its history.” This statement is not as broad as this question 

suggests. The facilities addressed are static in that they do not undergo major 

changes in design or fuel. The data given in the TBD, which address variability 

in workplace fields within these facilities, should therefore be reasonable relative 

to their uncertainties. 

(4) Page 64 (Burns and Mohrbacher 2004) – According to the Occupational External Dose 

TBD (Burns and Mohrbacher 2004), neutron track Type A (NTA) film and 

thermoluminescent dosimeters (TLDs) were only processed if the health physicist (HP} 

requested it, or if it was anticipated that the worker was exposed to neutrons. This seems 

to be somewhat circular reasoning, in that neutron dose was recorded only if it was 

known that neutrons were there. This policy could lead to missed neutron dose, 

especially in the early years before some radiation hazards were identified. Therefore, 

during DR, the worker could be assigned a missed dose based on limit of detection 

(LOD), instead of more realistic radiation worker’s doses. How will this shortcoming be 

addressed during neutron DR? 

The monitoring records of workers routinely exposed to neutrons would reflect 

this routine exposure. Additionally, neutron dose is assigned using 

neutron:photon ratios when it is apparent from the information provided that the 

employee may have been exposed to neutron radiation. 
 

 

 

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(5) Page 64 (Burns and Mohrbacher 2004) – NTA film track conversion factors are given as 

the number of tracks/cm2/mrem. Should this be in units of the number of 

tracks/mm2

-mrem? 

No. At 950× magnification, one field was 2 × 10-4 cm

2 

. 

(6) Page 65 (Burns and Mohrbacher 2004) – It states that as of 1969, ORNL began doubling 

the neutron dose values to obtain the recorded dose for unknown spectra. However, as 

stated elsewhere in the Occupational External Dose TBD (Burns and Mohrbacher 2004), 

some facilities had the total neutron spectra below the NTA film threshold. Doubling a 

zero recorded neutron dose (or only a small fraction of the total dose for some facilities) 

leads to a large missed dose. How will this be handled during DR? 

To the best of our knowledge, ORNL never assigned missed dose. The dose 

reconstructors typically assign dose using the approach in the TBD, using a 

neutron:gamma ratio. 

(7) Page 70 (Burns and Mohrbacher 2004) – It states that the dose reconstructor should use 

n-p values for neutron DR, unless it is known that the NTA film results were correct. 

Does this mean that all the discussions in the TBD (Burns and Mohrbacher 2004) 

concerning NTA film doses will not be used for the most part, but that the n-p method 

will be used for the majority of the neutron DRs? 

Correct. Neutron dose is assigned using a neutron:gamma ratio, which is the 

approach recommended in the TBD. 

(8) Page 71 (Burns and Mohrbacher 2004) – It states that the dose reconstructor should use 

the n-p dose ratios to estimate missed or unmonitored neutron doses before 1975. Is this 

the method proposed by the TBD (Burns and Mohrbacher 2004) for neutron DR that is to 

be used for unbadged workers who should have been monitored? How will the photon 

dose be assigned (LOD when photon values cannot be used for unmonitored workers 

because LOD only applies to workers wearing dosimeters)? Will additional technical 

information bulletins (OTIBs) be issued to cover neutron doses for unbadged workers? 

The TBD is referring to assigning dose for neutron workers. Unbadged workers 

should not have received any appreciable neutron dose. 

(9) Page 69 (Burns and Mohrbacher 2004) – TBD-6 recommends that the DR double the 

recorded neutron dose from 1950-present (2004) to account for different quality factors. 

Shouldn’t this only apply to NTA film and not TLDs? 

No. The TLD results still reflect NCRP 38 quality factors. 
 

 

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REFERENCES 

Burns, R. E., 2004a, Technical Basis Document for the Oak Ridge National Laboratory – 

Introduction, ORAUT-TKBS-0012-1, Rev. 00, Oak Ridge Associated Universities, Oak Ridge, 

Tennessee, August 11, 2004. 

Burns, R. E., 2004b, Technical Basis Document for the Oak Ridge National Laboratory – 

Occupational Environmental Dose, ORAUT-TKBS-0012-4, Rev. 00, Oak Ridge Associated 

Universities, Oak Ridge, Tennessee, May 7, 2004. 

Burns, Jr., R.E., and Mohrbacher, J.,2004. Technical Basis Document for the Oak Ridge 

National Laboratory – Occupational External Dose, ORAUT-TKBS-0012-6, Rev. 00; August 

11, 2004. Oak Ridge Associated Universities Team, Oak Ridge, Tennessee. 

Davis, D.M. and Gupton, E.D., 1957, Memorandum from D.M. Davis and E.D. Gupton to J.C. 

Hart, Radiation Dose in Diagnostic X-ray Procedures, March 26, 1957. 

Dickens, F., and Fleming, K., 2003. Technical Basis Document for the Oak Ridge National 

Laboratory – Site Description, ORAUT-TKBS-0012-2, Rev. 00; November 14, 2003. Oak 

Ridge Associated Universities Team, Oak Ridge, Tennessee. 

Gupton, E.D., and Tuck, T. L., 1956, Exposure Data Diagnostic Radiography, Oak Ridge 

National Laboratory, August 21, 1956. 

Lincoln, T.A. and Gupton, E.D., 1958, Radiation Dose to Gonads from Diagnostic X-ray 

Exposure, Journal of the American Medical Association, Vol. 166, January 18, 1958. 

NIOSH (National Institute for Occupational Safety and Health), 2002a, External Dose 

Reconstruction Implementation Guidelines, Revision 1, OCAS-IG-001, Office of Compensation 

Analysis and Support, Cincinnati, Ohio. 
 

 

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ATTACHMENT 4: SUMMARY OF CONFERENCE CALLS ON SC&A 

QUESTIONS PROVIDED TO NIOSH 

Introduction 

SC&A submitted written questions to NIOSH pertaining to the Oak Ridge National Laboratory 

Site Profile on August 1, 2006. NIOSH/ORAU provided written responses to these questions on 

August 25, 2006, in preparation for a conference call with SC&A. The questions and responses 

are provided in Attachment 3. Information provided by NIOSH/ORAU gives SC&A a more indepth knowledge of the rationale for assumptions made within the TBD and the source 

documents that provide the basis for the TBD. The summaries below are not verbatim 

discussion, but include information supplemental to the written responses provided by 

NIOSH/ORAU. They are arranged by general topics, since there was overlap in discussions 

conducted during the conference calls. The information provided by NIOSH/ORAU is listed 

under each topic. SC&A has provided comments towards the end of the summary. Action items 

resulting from discussions in the conference call are listed at the bottom of each summary. 

Participants 

NIOSH: John J. Johnson 

ORAU: Elyse Thomas, Robert Burns, Kenny Fleming, Liz Brackett 

Matt Smith, Mark Fishburn, Steve Cummings 

SC&A: Kathy Robertson-DeMers, Dr. Abe Zeitoun, Dr. Bob Bistline, Ron Buchanan, 

Dr. Harry Pettengill, Tom Bell 

ORNL Site Description TBD 

There was a wide range of particle sizes at ORNL. For particulate emissions from stacks, if they 

were larger particles, they would settle out on fallout trays. Smaller particulates would be 

captured by air monitors and included as a part of the data from these units. The height of the 

units is thought to be 6 feet above the ground level. The same air monitors were in place from 

1947 to the early 1980s. Ambient external dose for noble gases was measured initially with the 

use of ionization chambers (e.g., X-chambers). Inhalation of the particulates is addressed 

through the environmental data. The environmental dose assessment would be used for some 

who were not monitored. 

Discussions with the current Radiological Control Officer for the Life Sciences Division 

indicated that there had been no ruptured sources at the ORNL facilities at the Y-12 Plant. 

Sealed sources were used for animal irradiation, and did not pose an internal hazard. A form was 

filled out with the use of each source. The focus on sources in this particular area of the TBD 

was with the Biology Division. 

We have tried to pick up information about ruptured sources during the worker interviews and 

the CATI interviews, but have not had any reports of this happening. The Radiation Area 

supervisor was contacted and provided NIOSH/ORAU with a list of sources handled in the Life 

Sciences Division. From the information that has been gained, there appear to have never been 
 

 

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any ruptured sources. NIOSH/ORAU relies primarily on the CATIs and the information 

obtained from Worker Outreach meetings. They do not generally try to directly contact specific 

workers, but did contact many individuals within the Radiological Control Operation to obtain 

additional information. 

There are many dates of specific operations and processes that were not included within the 

ORNL Site Profile. Due to the nature of research that took place at the site, individual log books 

were maintained by radiation protection staff. These log books, located in storage, denote daily 

and shift-wise tasks that took place within radiological areas. Where MDA data is missing, the 

dose reconstructor is provided with claimant-favorable assumptions (e.g., energy ranges, time 

spent in radiological areas, etc.) to reconstruct dose. For monitored workers, bioassay data are 

used to assess individual exposures. For unmonitored workers or workers that have gaps in their 

data, co-worker data are used. 

SC&A Conference Call Comments 

SC&A is concerned not only about the possibility of inhalation for unmonitored workers, but 

also for monitored workers. Some consideration should be given to the particle size used to 

assess these releases. At other sites, particle deposition on the skin has also resulted in some 

significant doses to personnel. These items deserve further consideration. 

The presence of so many processes and operations at ORNL, which introduce numerous 

radionuclides, is of concern. The Site Description TBD (Fleming 2006b) lists radionuclides and 

processes that have not been considered in the Internal Dose TBD (Bollenbacher et al. 2006). 

Table 5A-2 (page 39) shows several time periods where MDA data is missing. It is unclear how 

dose assignments will be handled for these time periods, or if there was a potential for exposure. 

Action Items for NIOSH/ORAU 

None. 

Action Items for SC&A 

(1) Provide a list of action items from the conference call. 

(2) Provide a summary of the conference call discussion. 

Occupational Medical Dose (ORAUT-TKBS-0012-3) 

There is no attempt to subtract any other non-occupational x-ray exposures from the dose of 

record. NIOSH/ORAU generally do not have enough data to ascertain whether individuals wore 

dosimeters during x-ray procedures unless the CATI alerts us to this possibility. Correspondence 

from an area physician indicated a concern regarding some workers being allowed to stand inside 

the x-ray room when other individuals were receiving chest x-rays. This was done because there 

were no changing room facilities outside the examination rooms. To resolve the concern, ORNL 

placed thermoluminescent dosimeters in locations within the x-ray room for a month where 

workers may have spent time during x-ray procedures to document low exposures. The study 
 

 

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indicated that potential exposures from scattered radiation would not have exceeded 0.08 mrad 

per chest x-ray examination. 

X-ray exams that were required as a “condition of employment” are included in dose 

assessments. NIOSH/ORAU agrees that screening exams done during employment does 

constitute an occupational medical exposure. This is inclusive of asymptomatic as well as 

symptomatic people involved in medical screening. 

Medical files (including records of x-ray examinations) are available dating back to the 

beginning of site operations for ORNL employees, and are stored in the ORNL Medical Records 

vault. It appears that records from medical examinations conducted at the Oak Ridge Hospital 

were transferred to the ORNL vault, although there is some uncertainty as to whether all 

photofluorography (PFG) films completed at the hospital were transferred. The medical records 

that were reviewed indicated dates that x-ray examinations were conducted. The Internal Dose 

TBD directs the dose reconstructor to assume all pre-employment chest x-ray examinations were 

PFGs, and to assume that all other routine or re-takes used 14” x 17” films. 

ORAU, with the support of ORNL medical staff, have reviewed a portion of the large set of 

x-ray records (including films) at the ORNL records vault. The review of a subset of PFG film 

records indicated they were concentrated around the 1943–1944 timeframe. An individual, from 

the reviewed set of records, was found to have had two PFGs about a year apart. A discussion 

with the individual that began working at ORNL as the Radiology (x-ray) Technician in 1947 

indicated that chest PFGs were all conducted at Oak Ridge Hospital, and would have been 

conducted prior to the ORNL x-ray facilities commencing operations (October 1947). PFGs 

were not observed in any individual medical jackets, but were stored on six shelves with other 

PFGs that were in the vault. Though no PFG records after 1944 were observed in the vault, it is 

possible that some may exist, but not very probable. There were several hundred PFGs located 

on the shelves, and approximately 100 were observed during the site visit. 

In the mid-1950s, Dr. K.Z. Morgan instituted a study looking at x-ray exposures from both 

ORNL Medical and area radiology clinics/hospitals. The results of the study are documented in 

two peer-reviewed papers written by Drs. Lincoln and Gupton. NIOSH agrees that the purpose 

of the study was to assess, and possibly determine a mechanism to reduce, gonad exposures from 

x-ray exams that individuals had at ORNL and offsite. 

The filtration used in the Picker Model R-2 unit installed in October 1947 was documented in 

each of the papers referred to in the preceding paragraph to be 1.0 mm Al total filtration for 

estimation of entrance skin exposure (ESE). A value of 0.04 mm Al inherent filtration was 

provided by the Radiology Technician, but assumed to be in error, and a half value layer of 

1.5 mm Al was actually used to estimate ESEs. Conversations with Ron Kathren and others 

indicated that 0.04 mm Al filtration would not have taken into account all the shielding that 

would have been present in the x-ray tube and housing. As a result, alternate filtration values 

were used to estimate ESEs. The Radiology Technician pointed out aluminum wedges in a box 

during a site visit as being used to harden the beam for lumbar spine exams. 
 

 

 

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Discussions with a Radiology Technician at ORNL from 1947 indicated that the beam was well 

collimated. There was also a 20-cm cone collimator documented in the referenced papers as 

being used at least from the mid-1950s. 

ORNL provided NIOSH/ORAU with a page of text that identified the equipment that is and was 

used at ORNL and the operating characteristics. The text provided by ORNL was not in a 

consistent format and, therefore, the information was extracted from the text and re-organized in 

Table 3-2. Operating parameters, and the 0.04 mm filtration, were cited in an early handwritten 

survey conducted to support the study referred to above. The record of the survey is on the 

NIOSH/ORAU site research database. However, after reviewing and comparing the operating 

parameters and instruments in use at the time, the 0.04 mm filtration value and exposure time for 

the Westinghouse Riviera unit were changed. The assumed half value layer was discussed in a 

previous paragraph. The exposure time was increased by a factor of 10, because the provided 

exposure appeared low. This assumption is favorable to the claimant. Other than those two 

items, NIOSH/ORAU relied entirely on information supplied by the ORNL medical staff. 

Based upon the reviews of folders made during the visits to the site, NIOSH/ORAU believe the 

medical files exist for all individuals that have been and still are employed at ORNL (except 

those that may have been transferred to other sites within the Oak Ridge Operations). These 

medical files contain color-coded sheets that indicate the date and type of x-ray examination 

conducted. This available and provided information can be used by dose reconstructors to 

reconstruct exposure that are favorable to claimants. NIOSH/ORAU believe the medical files 

contain all necessary information on medical x-ray exposures needed to estimate energy 

employee dose. This includes the Radiology report for particular x-rays. 

There has been research into the possibility of workers lining up inside the x-ray room when 

chest PFG screenings were being conducted. It is not known what the conditions were in the 

Oak Ridge Hospital, or what was done at the hospital to ensure that people stepped out of the 

primary beam area in the x-ray room. In addition to other assumptions that ensure dose 

reconstructions are conducted that are favorable to claimants, 30% uncertainty is added to the 

occupational medical x-ray dose. NIOSH/ORAU is conservative on the estimation of ESE, and 

is confident that dose reconstructions are favorable to claimants. The exposure to individuals 

standing in the room would primarily be from scatter. This dose would not be significant. 

SC&A Conference Call Comments 

The responses provided to our questions were fairly straightforward. There are a 

few items that require further clarification. 

There was a period of time (prior to October 3, 1947) when pre-placement chest PFG 

examinations were conducted at the Oak Ridge Hospital. The TBD states that screening 

was done at the Oak Ridge Hospital, and that “no PFGs exams of the chest were 

performed at the ORNL site during physical examinations;” however, PFG films and 

related records were transferred to ORNL Medical. It appears that at a later date, the 

PFG films were transferred to ORNL. Use of PFGs for tuberculosis screening was a 

general application used up to the late 1950s. There should be some level of 
 

 

 

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investigation into whether ORNL workers continued to be sent to the Oak Ridge Hospital 

(or any other medical facility) for PFGs after 1947, including for tuberculosis screening. 

A call was made to Tim Vinson at Methodist Hospital in 2003 to inquire about the type 

of PFG equipment used in the mid-1940s, and whether PFG films may still be available 

at the hospital. He thought this was not likely. SC&A believes NIOSH should 

investigate this further beyond the interview conducted. 

NIOSH/ORAU relied on responses from one medical staff person at ORNL, and a review of 

those responses by the Medical Director, on the frequency of PA chest, LAT chest, and lumbar 

spine films. Further review of documentation and interviews with former medical directors who 

actually set policy is encouraged. 

In a letter from Dr. T.A. Lincoln to Dr. E. G. Ammermann of the University of Tennessee 

Memorial Hospital, Dr. Lincoln of ORNL stated the following: 

Seaton Garrett told me about his recent conversation with you regarding x-ray 

exposures in the Health Division of the Oak Ridge National Laboratory. He 

indicated that you were quite critical of our practice of allowing other patients to 

be in the room at the same time a patient gets his periodic chest x-ray. 

I want to reassure you that we, too, are concerned about any radiation exposure 

our employees get. (Lincoln 1970) 

Dr. Lincoln agrees later in his letter that patients should not be in the examining room when 

another patient is getting an x-ray; however, in order to accomplish the number of exams 

required, a common practice allowed other workers in the exam room at the same time. This 

clearly indicates that individuals were allowed in the room while others were receiving x-rays. 

ORAU, in their response to SC&A questions on August 25, 2006, indicated that the extra 

exposure to individuals waiting in the x-ray room is likely not greater than the 1% extra dose 

estimated by Lincoln. We agree that the total dose from this type of scatter exposure is not likely 

to be significant. It was estimated by Dr. Lincoln that the “exposure was probably less than 

1 mrad.” (Lincoln 1970). This practice, if it occurred frequently, and how it adds to the 

occupational medical dose should be considered. 

The use of a 1 mm Al filtration is questionable on the Picker Model R-2. This may not 

be accurate. 

The Occupational Medical Dose TBD (Fleming 2004a, pg. 7) indicated the use of the 

following assumption. 

Given the disparity between this value and what is typical, a beam quality with an 

HVL of 1.5 mm Al was assumed for the assessment of worker exposures from this 

unit. The total filtration given for this device was found documented (Lincoln and 

Gupton 1958a and 1958b) as 1 mm Al and was used to estimate the entrance skin 

exposure2 (ESE) values for lumbar spine X-ray examinations for the instrument 

used between 1947 to 1963. The documented skin exposure in these same reports 
 

 

 

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was low (21 mrad) for the PA chest X-ray examination as compared to skin 

exposures from other chest X rays given during that timeframe. The skin exposure 

for that timeframe was estimated using operating parameters. 

Although the TBD appears conservative, the filtration is likely greater than this and probably on 

the order of 2.5 to 2.75 mm Al. 

Even though the TBD suggests collimation was always used at ORNL, there is no verifiable 

evidence to suggest the 20-cm cones were effective. ORAU derived organ doses based on an 

assumption of a poorly collimated beam (Fleming 2006a, pg. 10). Furthermore, efforts should be 

made to verify the size of x-ray unit cones and when they were used. 

Action Items for NIOSH/ORAU 

None. 

Action Items for SC&A 

Provide NIOSH/ORAU with a copy of Lincoln, T.A., 1970, Correspondence between 

T.A. Lincoln and E. G. Ammermann dated September 15, 1970, no subject, Oak Ridge 

National Laboratory, Oak Ridge, Tennessee. 

(Provided via fax on September 6, 2006) 

Occupational Environmental Dose (ORAUT-TKBS-0012-4) 

The Environmental Dose TBD focuses on unmonitored workers. Monitored workers would not 

be within the scope of this TBD. The dose reconstructor relies on ORAUT-PROC-0060, 

Occupational Onsite Ambient Dose Reconstruction for DOE Sites (Winslow 2006a) and 

ORAUT-PROC-0061, Occupational X-ray Dose Reconstruction for DOE Sites (Winslow 2006b) 

for assignment of environmental and medical dose, respectively. These procedures describe how 

to address dose from episodic releases. 

In the absence of data from the site, NIOSH/ORAU makes use of studies previously completed 

by other groups, such as the Oak Ridge Dose Reconstruction project. The TBD assumes 

conservative dispersion coefficients. Particular radionuclides yielding the highest organ dose are 

assumed for the gross alpha counts prior to the 1950s. 

The environmental dose for ORNL individuals located at the Y-12 Plant is based on the higher of 

the doses between any facilities in Oak Ridge where an individual worked. Employment 

determinations are made from information provided in NOCTS. If individuals worked at all 

three sites, the values for K-25 also would be considered. 

NIOSH/ORAU feels that dose from resuspended contaminated soil would be captured in the air 

monitoring data. 
 

 

 

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SC&A Conference Call Comments 

The Environmental Dose TBD overall is well done. 

There is a concern that episodic releases (accidents) have not been considered where exposure 

may be acute rather than chronic. The annual average value may not provide a bounding value 

for accident scenarios. Some of the releases at ORNL were between 2 and 24 hours. These 

accidents are mentioned in the Site Description TBD, but not in the Environmental Dose TBD. 

In these cases, there is a lot of uncertainty about dose to unmonitored workers from outside the 

normal facilities operations where the worker might have received dose. 

The TBD has not taken into consideration any environmental dose other than airborne releases. 

Certain exposures to radioactive waste may have taken place. There are waste disposition areas, 

such as the north and south ponds, which may add to exposures. Resuspension and ingestion of 

radioactive contaminants ought to be considered when dose is reconstructed. 

ORNL was involved with a large number of radionuclides, which could have caused exposure 

during accidents and routine releases. When there is a release from a reactor, there is a whole 

array of chemicals and radioisotopes released. These radionuclides should be considered in the 

environmental dose analysis in order to ensure a bounding estimate. 

Action Items for NIOSH/ORAU 

NIOSH/ORAU will review the differences between the radionuclides in the Site Description and 

what is apparent from air monitoring data. 

Action Items for SC&A 

None. 

Occupational Internal Dose (ORAUT-TKBS-0012-5) 

When dealing with unmonitored workers, NIOSH/ORAU uses co-worker data, if the workers 

location was unknown and they may have been exposed. If they were known to work in an area 

with potential exposure, but were not monitored, a co-worker dose is also used. 

Table 5-3 of the Internal Dose TBD was obtained directly from ORNL documents, including the 

footnotes. The reference to consulting the Internal Dose Group was guidance for ORNL 

personnel, rather than for NIOSH/ORAU personnel. Internal dose estimates are based on 

guidance provided in ORAUT-OTIB-0018 (Brackett and Bihl 2005) and ORAUT-OTIB-0019 

(Brackett 2004). Table 5-3 includes only the most abundant radionuclides. 

NIOSH/ORAU is not aware of the use of tritides at ORNL. Building 7025 at the east end of the 

site was involved in tritium work (i.e., target fabrication), but tritides were not involved. The 

existence of tritides at ORNL has not been investigated. 
 

 

 

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NIOSH/ORAU is currently using a revised IMBA program specifically designed for use in the 

compensation program. The IMBA program has no information related to solubilities. IMBA 

(ORAU) has some default selections for uranium enrichment; however, the dose reconstrctor 

assumes 100% U-234, which has the largest dose coefficients. This gives all workers exposed to 

uranium the benefit of the doubt. There is no default solubility class for uranium or other 

radionuclides. All cases are done so that the most claimant-favorable solubility class is assumed. 

Background uranium in bioassay samples is not subtracted from the gross measurement. 

All uranium is assumed to be from occupational exposure, which is claimant favorable. 

Radon exposure would be included by the dose reconstructor where exposure to radon was 

indicated. Currently, radon dose is not being looked into at ORNL. The assessment of radon 

dose would apply only to lung cancers, and for the early period, claimants with lung cancer are 

being compensated. Therefore, there is probably no impact for not assessing radon dose. 

It would be helpful to have a discussion about when to assume Am-241 exposure and when to 

assume Cm-244 exposure. Some information on the timeframe and possible location for 

curium-244 in the TBD would be helpful. The use of Am-241 versus Cm-244 is based on the 

purity and time period it was handled. 

The dose reconstructor will determine if the individual has a potential for exposure and apply the 

co-worker dose or the environmental dose. Chronic intakes are assigned to individuals with 

bioassay data. 

There is not much difference between inhalation and ingestion dose. Inhalation is assumed as 

the mode of intake, which is a conservative assumption. In the case of soluble compounds, there 

is little difference between the dose from inhalation and ingestion. When using air monitoring 

data, there is a fraction added for ingestion. 

The Super Type S OTIB will outline how to assess doses from intakes of this type. This is a 

general document; however, some site-specific documentation will have to be given. Any site 

working with PuO2 or Pu will need to be evaluated for the possibility of exposure to high-fired 

oxides. The model has been developed; however, the team is in the process of identifying when 

it will be applied. 

The MDA values for ruthenium were not questioned. It is somewhat strange that the MDA for 

ruthenium increases rather than decreases over time. 

There is no plan to revise the ORNL TBDs further at the current time. 

SC&A Conference Call Comments 

There is a general concern over the lack of consideration for exposure to a number of 

radionuclides. Dose from the ingrowth of Am-241 has not been considered. Tritides, which are 

not picked up by traditional monitoring methods, have not been investigated. The defaulting to 

Am-241 as the transplutonium radionuclide of choice, when Cm-244 was handled at the site, 
 

 

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potentially leads to missed dose. In a June 1944 memo, it was determined that radon levels were 

as high as 6,700 pCi/L, which constituted a significant dose hazard. With the handling of radium 

in the early years, radon exposure should be evaluated for those workers. There may be 

processes at ORNL involving exposure to high-fired plutonium oxide, such as with the fuel 

cycle, microspheres and gels, and catalytic carbon, where this will be involved. 

There is no mention in the TBD regarding dose from ingestion. With some of the highly soluble 

isotopes used at ORNL, there is a possibility of workers receiving ingestion dose. Ingestion 

would be particularly relevant to the assignment of dose using air-monitoring data in the 1940s. 

Some of the minimum detectable activities (MDAs) on page 16 and Table 5-9 (page 39) in 

Appendix 5A have wide variabilities. For example, Ru-106 has an MDA in 1951 of 

0.03 dpm/24 hour sample. It is counterintuitive to believe that as equipment and methods 

improved, the MDA would increase. By 1960, the MDA jumps to 61.85 dpm/24-hours. 

Inconsistencies exist between the ORNL and LANL TBDs with regard to recommendations for 

the default solubility class for uranium. 

Action Items for NIOSH/ORAU 

(1) NIOSH/ORAU will investigate the potential for occupational exposure to radon at 

ORNL. 

(2) NIOSH/ORAU will evaluate whether high-fired oxides and tritides are potential 

sources of exposure at ORNL. 

(3) The inclusion of dose from Am-241, as an impurity in plutonium, will be included 

in subsequent revisions of the TBD. 

Action Items for SC&A 

None. 

Occupation External Dose (ORAUT-TBKS-0012-6) 

Workers entering restricted areas more than 3 days per week were assigned routine beta-gamma 

monitors. Those entering the areas for less than 3 days a week were not assigned a routine 

dosimeter, but they were provided with a temporary dosimeter. PICs were assigned to these 

employees, like any other employee. 

On the handwritten records, the blank indicates that an individual was not monitored. A zero 

indicates that the dose was less than the minimum detectable dose. Both types of records are 

provided in claimant files. Between the manual cards and the computer printouts, the dose 

reconstructor will use the form that provides the most information. This is usually the hardcopy 

record. 
 

 

 

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The skin dose will be assigned based on ORAUT-OTIB-0017 (Merwin 2005), which addresses 

assigning dose to the skin, breast, and testes. There is skin dose data prior to 1961, which can be 

used to assign dose. No reliable data during the period of time when PICs were used is available. 

A co-worker dose can be used to assign dose where necessary. 

There was also enough data (experimental in nature) in the early period of 1943–1944 to develop 

a co-worker model for 1943–1944. This data was likely film badge data. The co-worker model 

was developed from CEDR data which was obtained from ORNL. ORAUT-OTIB-0021 

(Merwin 2004) provides the details of the model. Unmonitored workers would receive this 

appropriate co-worker dose. 

The primary basis for assigning neutron dose in the NTA era and before is the application of a 

neutron-to-photon (n-p) ratio. It is felt that the later spectral is representative of conditions in the 

earlier years. NTA results are only used when the NTA dose exceeds that calculated with the n-p 

ratio. In the mid- to late-1940s, operations at the site primarily involved the Graphite Reactor. 

The neutrons encountered were mostly thermal neutrons. There is no reason to suspect that 

changes would occur in the n/p for the Graphite Reactor. 

The fraction of dose equivalent above the energy cutoff for NTA film will not be used for dose 

reconstruction. 

Missed neutron dose is applied to those that are monitored during their career. If the worker 

indicates through the CATI that they were exposed to neutrons, the n/p ratio is assigned to the 

co-worker doses. The n-p ratio will be used to determine doses for those individuals who 

received neutron exposure. In OTIB-0021, Section 4, paragraph 4 (Merwin 2004), it does 

discuss neutron dose and it recommends use of co-worker data for unmonitored workers. 

Applying n-p ratios in cases where neutron exposure is expected is not covered in OTIB-0021. 

The TBD predates the OTIB. 

The correction factor of 2 should be applied to NTA film, as well as TLDs. The correction factor 

converts the site-derived dose based on NCRP 38 (NCRP 1971) to a dose based on ICRP 60 

(ICRP 1990). This factor has been applied at almost all DOE sites. 

SC&A Conference Call Comments 

The TBD requires some clarification on how external monitoring was determined, and how 

comprehensive this monitoring was. This is especially true of the early years. Clarification is 

also needed on whether PICs were used in lieu of film badges, especially in the 1940s. 

There needs to be further verification that the n-p ratios in the 1980s are representative of those 

in the 1940s. The equivalency of the facilities has not been demonstrated. 
 

 

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Action Items for NIOSH/ORAU 

(1) NIOSH/ORAU will investigate whether the comment under the table in Attachment 6C is 

appropriate. 

Action Items for SC&A 

None. 

References 

Brackett, E. M., 2004, Technical Information Bulletin – Analysis of Coworker Bioassay Data for 

Internal Dose Assignment, ORAUT-OTIB-0019, Rev. 00, Oak Ridge Associated Universities, 

Oak Ridge, Tennessee, December 29, 2004. 

Brackett, E.M., and Bihl, D.E., 2005, Internal Dose Overestimates for Facilities with Air 

Sampling Programs, ORAUT-OTIB-0018, Revision 01, Oak Ridge Associated Universities, Oak 

Ridge, Tennessee, August 9, 2005. 
 

 

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ATTACHMENT 5: CONSISTENCY AMONG SITE PROFILES 

The default site profile assumptions and methodologies for Oak Ridge Associated Universities 

(ONRL) are summarized below and were compared to those of other site profiles reviewed to 

date or in current review. Site profiles completed to date by the SC&A team include Bethlehem 

Steel, Mallinckrodt Chemical Works (MCW), Iowa Army Ammunition Plant (IAAP), Hanford, 

the Savannah River Site (SRS), the Y-12 National Security Complex (Y-12 Plant), Idaho 

National Engineering and Environmental Laboratory (INEEL), the Nevada Test Site (NTS), the 

Rocky Flats Plant (RFP), the Mound Plant (Mound), Los Alamos National Laboratory (LANL), 

the Linde Ceramics Plant (Linde), and the Pinellas Plant. Additional site profiles in the process 

of review are the Fernald Environmental Management Project (Fernald) and the Paducah 

Gaseous Diffusion Plant (Paducah). ORNL had multiple missions that overlap with a number of 

other sites in the DOE complex, such as weapons research, the heat source program, reactor 

research, assembly and disassembly operations, and tritium operations, to name a few. 

To ascertain the differences in assumptions between what assumptions are used for the ORNL 

Site Profile versus other site profiles, the assumptions from each ORNL TBD must first be 

understood. The core assumptions for each TBD have been outlined below. 

Dose Reconstruction Assumptions for Occupational Medical Exposure 

Diagnostic x-ray procedures required as a “condition of employment” included preplacement, 

routine annual, and termination exams. There was a limited amount of site-specific data related 

to x-ray equipment and techniques in the TBD. Table 3-3 in the Occupational Medical Dose 

TBD (Fleming 2006a, pg. 18) summarized the type of x-ray equipment used at ORNL from 1947 

to 2002, and provided information on location, techniques, x-ray conditions, type of people 

receiving examinations, and the changing frequency of chest x-rays, depending on age, after 

1990. Information provided in ORAUT-TKBS-0012-3, Table 3-2 (Fleming 2006a, pg. 17) 

indicates that x-ray frequency through the operating period ranged from annually to every 

3 years and after 1990, from annually to every 2 years, depending upon age. Table 3-3 (Fleming 

2006a, pg. 18) provides ORNL-provided operating parameters and assumed operating 

parameters for the purposes of dose reconstruction. Claimant-specific information is provided to 

NIOSH by ORNL, and is used to supplement the frequencies listed in ORAUT-TKBS-0012-3, 

Table 3-2 (Fleming 2006a, pg. 17). The basis for the ORNL Occupational Medical Dose TBD 

(Fleming 2006a) is the External Dose Reconstruction Implementation Guideline (NIOSH 2002) 

and ORAUT-OTIB-0006, Technical Information Bulletin: Dose Reconstruction from 

Occupationally Related Diagnostic X-ray Procedures, Revision 2 (Kathren 2003b). In the 

absence of site-specific data, the ORNL Site Profile relied on values provided in this document 

for occupational medical exposure. 

The default values for occupational medical dose are the same as those used in other site profiles, 

as noted on the following pages of the Occupational Medical Dose TBD (Fleming 2006a). 

• 	 Routine x-ray frequency through the operating period ranged from annually to every 

3 years. Claimant-specific information was used to supplement the frequencies listed in 

ORAUT-TKBS-0012-3, Table 3-2 (pg. 17). 
 

 

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• 	 Exam included posterior-anterior (PA) chest x-rays, lateral chest x-rays, anteriorposterior (AP) lumbar spine x-rays, lateral lumbar spine x-rays, and stereoscopic 

photofluorography (PFG) with two views (pg. 20). 

• 	 All employees were assumed to have received a preplacement stereo exam if employed 

between 1943 and October 3, 1947. Lumbar spine exams were assumed for crafts 

personnel from April 6, 1950, through September 23, 1953 (pg. 18). 

• 	 PFG dose values provided in Table 3-4 assume two exposures. This value should be 

divided by two if the record indicates only a single view (pg. 11). 

• 	 The PA chest and AP Lumbar spine thicknesses were assumed to be 26 cm. Thickness 

used for lateral lumbar spine and lateral chest x-rays was 34 cm. A 5-cm distance was 

assumed between the film and the individual’s body. Adjustments were made for larger 

workers. For example, lateral chest x-rays were estimated for large workers by 

multiplying the PA Entrance Skin Exposure (ESE) by a factor of 2.0 (pp. 9–10). 

• 	 Dose conversion factors (DCFs) prior to 1963 were based on poor collimation and thus 

were obtained from ORAUT-OTIB-0006, Table 4.0-1 (ORAU 2003b). After 1963, the 

units were considered to be well collimated, and DCFs outlined in International 

Commission on Radiation Protection (ICRP) Publication 34 (ICRP 1982) were used 

(pp. 15, 20–21). 

• 	 Source-to-image distance for chest, PFG chest, and lumbar spine exams were assumed to 

be 183 cm, 122 cm, and 99 cm, respectively (pg. 18). 

• 	 The ESE for the AP lumbar spine x-ray presents both the AP and spot AP exposures. 

• 	 The ESE for the lateral lumbar spine x-ray presents both the lateral and spot lateral 

exposures. 

• 	 Single-phase units were assumed until 1990 (pg. 10). 

• 	 Values are dimensionless backscatter factors from Table B.8 of NCRP 102 (NCRP 1989). 

Values for half value layers = 2.5 and 3.5 were obtained via linear interpolation. (pg. 22, 

footnote to Table 3.5). 

• 	 Skin dose calculations include backscatter factors from Table B.8 in NCRP 102 (NCRP 

1989). Values for half value layers of 2.5 and 3.5 were determined by linear interpolation 

(pp. 21 and 23). 

• 	 For organs included in IREP that are not specifically identified in ICRP 34, the DCFs for 

analogue organs that were anatomically the closest to the organ in question were used. 

(pp. 14 and 34). 
 

 

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• 	 Table 3-2 does show techniques and x-ray conditions for several different types of x-ray 

equipment used from 1947 to 2002 (pg. 26). 

• 	 DCFs for lumbar spine examinations are not given in ICRP 34; therefore, values for the 

upper gastrointestinal tract exams (AP and LAT) were used (Table 3-5, pp. 20–21). 

• 	 Uncertainty for x-ray procedures is 30% (pg. 11). 

Overall, the default values assigned for determining medical exposure are relatively consistent 

among site profiles. The site profile applies OTIB-0006, Revision 2, dated December 29, 2003, 

although a subsequent version was released on December 21, 2005. This version predates the 

issuance of Rev. 01 PC-1 of the TBD, dated July 21, 2006. The later version of OTIB-0006 

(Kathren and Shockley 2005) provides guidance on the assignment of dose from lumbar spin 

exams, which were given at ORNL. This should be corrected in subsequent revisions of the 

TBDs. Other deviations from the standard assumptions are based on site-specific information. 

Dose Reconstruction Assumptions for Occupational Environmental Exposure 

ORAUT-TKBS-0012-4 (Burns 2004b) describes the default assumptions for occupational 

environmental dose at ORNL. Occupational environmental dose included internal exposures 

from onsite atmospheric radionuclide concentrations, and external exposure from submersion 

and ambient radiation. Values derived for occupational environmental dose were based on 

environmental measurements (i.e., air monitoring data) and emission rates. Attachment 4B 

(pg. 40) provides a summary of the annual average airborne concentration data and the 

associated intakes for an individual assumed to be exposed to the average concentration for 

2,000 hours breathing at a rate of 1.2 m3

 per hour. 

Other assumptions made with respect to environmental dose from the Occupational 

Environmental Dose TBD (Burns 2004b) include the following: 

• 	 “The three principal contributors to inhalation dose individuals could have been exposed 

to in the outdoor air on the ORNL site were I -131, particulate mixed fission products 

(MFPs), and tritium. Lack of available monitoring data, for I-131 in particular, required 

that bounding values of dispersion coefficients coupled with available source term 

estimations be used to establish the needed onsite airborne concentration and intake data 

for certain periods” (Burns 2004b, pg. 27). 

• 	 All airborne concentration data are provided in terms of annual averages (pg. 27). 

• 	 Site-wide annual average exposure rate values include natural background and 

contributions from fallout (pg. 27). 

• 	 Annual average site-wide exposure rates for 1947 were assigned to 1944 through 1946. 

The average exposure rate for 1970 was assigned to years 1971 through 1975. The 1984 

data was assumed for 1985–2003, and 1982 (pg. 27). 
 

 

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• 	 Intakes associated with annual average airborne concentration data were based on 

2,000 hours of exposure with a breathing rate of 1.2 m3/hour (pg. 28). 

• 	 Values developed by ORNL were used to estimate ground-level concentrations resulting 

from local, low-elevation releases (pg. 28). 

• 	 From 1944 through 1965, airborne I-131 concentrations on the ORNL site had to be 

established using information on emission rates and bounding values of dispersion 

coefficients. Three different dispersion coefficients were used to derive local onsite 

concentrations of I-131 from the 3020 stack, the 3039 stack, and fugitive emissions 

(pg. 29). 

• 	 The total quantity of I-131 available from the processes was used to determine bounding, 

ground level concentrations until 1961 (pg. 30). 

• 	 A factor of 2 is applied to I-131 release data in 1961 through 1965 to account for sample 

line losses (pg. 31). 

• 	 The average values measured from the charcoal cartridges for the Local Air Monitoring 

(LAM) stations were used to determine concentrations for 1966–1984. The average 

annual concentration from 1984 was applied to subsequent years (pg. 31). 

• 	 I-131 is considered a soluble or reactive vapor (pg. 31). 

• 	 Annual concentration and intake data are provided for MFPs. These values are based on 

the average reported gross beta measurements from the LAM network. MFP dose is 

considered negligible; however, if included in the dose reconstruction, the recommended 

activity should be based on Type S Ru-106 or Ce-144, depending on which yields a 

higher dose for the organ of interest (pg. 31). 

• 	 No monitoring data were available for airborne tritium concentrations; therefore, 

concentrations were established based on inventory, release data and conservative 

diffusion coefficients prior to 1984. Tritium is not considered a significant source term 

before 1967 (pp. 32, 39–40). 

• 	 Intake values for tritium include a factor of 1.5 to account for direct skin exposure. The 

assumed form of tritium is HTO (tritiated water vapor). A geometric standard deviation 

of three is used (pg. 32). 

• 	 A site-wide average was chosen to be the most representative of the unmonitored 

worker’s exposure profile. If the workers dose is identified as coming from a specific 

source, the values for the specific source or the site wide-average can be chosen (pg. 33). 

• 	 A summary of total annual airborne releases of I-131, tritium, Kr-85, and Xe-133 

reported by ORNL is included in Attachment 4A. All values are given in total curies 

discharged to the atmosphere (pg. 39). 
 

 

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Review of site profiles to date indicates that the NIOSH/ORAU team has not come to a 

consensus on what components should be considered in the environmental dose. Dose from the 

resuspension of contaminated soil has not been considered in the Occupational Environmental 

Dose TBD (Burns 2004b). This source of exposure was considered at the Rocky Flats Plant 

(McDowell-Boyer and Little 2004) and the Savannah River Site (Scalsky 2005). Ambient 

external exposure from contaminated soil was also not considered. As with many of the site 

profiles, there is no consideration of potential exposure from liquid effluents at ORNL. 

Dose Reconstruction Assumptions for Occupational Internal Exposure 

The internal dose TBD, ORAUT-TKBS-0012-5 (Bollenbacher et al. 2006), describes the default 

assumptions for occupational internal dose at ORNL. The assumptions were derived from 

historical records relating to the in-vivo, the in-vitro, and the air monitoring programs. As shown 

in Table 5A-2 on page 39 of the Internal Dose TBD (Bollenbacher et al. 2006), MDAs have been 

determined for gross alpha, gross beta, and 16 radionuclides found in urinalysis sampling 

(Am-241, Cm-244, Cs-137, H-3, I-131, Np-237, Pm-147, plutonium, polonium, Pu-241, Ra-226, 

rare earths, Ru-106, Sr-89 + Sr-90, and gross alpha), and 4 radionuclides found in fecal samples 

(Am-241, Cm-244, plutonium, and Th-232). Figures 5A-1 through 5A-18 (Bollenbacher et al. 

2006, pp. 40–65) show plots of data obtained and calculated for the following radioisotopes: 

(urine) Am-241, Cu-244, Cs-137, gross alpha, gross beta, H-3, Pm-147, Pu-238, Pu-239, Pu-230, 

rare earths, Sr-90, U-233, U-238; and (feces) Am-241, gross alpha, Pu-239, Th-232. In the first 

plot, the MDA values were plotted against the date of their analysis. 

Other assumptions related to in-vitro analysis in the Occupational Internal Dose TBD 

(Bollenbacher et al. 2006) include the following: 

• 	 Because of the diverse operations occurring at ORNL, the compilation of a 

comprehensive list of radionuclides handled by ORNL employees would be difficult to 

assemble. The radionuclides assumed to produce a measurable dose at this facility 

include uranium, activation products, fission products, and transuranics (pg. 10). 

• 	 Dose reconstructors should choose the radionuclide and tissue(s) of interest using the 

solubility classes in ICRP Publication 66 (ICRP 1994). (pg. 10). 

• 	 Assume inhalation with a particle size of 5-micron Activity Median Aerodynamic 

Diameter (AMAD) particle size (pg. 10). 

• 	 Table 5-2, page 10, provides the assumptions on solubility classifications used by ORNL 

for re-evaluation of historical bioassay results. The default absorption for uranium is 

Type S. 

• 	 In-vitro monitoring data from an ORNL database were used to estimate MDAs. The 

MDAs were calculated from analytical records recovered from the database (pg. 12). 
 

 

 

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• 	 Plutonium-238, Pu-239, and Pu-240 were carried through in the total plutonium analysis. 

Alpha spectroscopy in 1989 allowed the lab to distinguish between Pu-238 and Pu-239. 

Plutonium-241 is reported separately, because it is a beta emitter (pg. 15). 

• 	 Table 5-9 provided the annual average MDAs (dpm/24-hr sample) for 1947 through 

1989, with the blanks indicating no analytical results were found (pg. 16). 

• 	 Assume Am-241 for measurements involving trivalent alpha actinides, unless results 

suggest otherwise (pg. 18). 

• 	 The default radionuclide assumed for positive total plutonium measurements is Pu-238 

(pg. 18). 

• 	 There is no mention of subtracting background radiation from the in-vitro results in the 

ORNL Internal Dose TBD. 

Other assumptions discussed in the ORNL Occupational Internal Doe TBD (Bollenbacher et al. 

2006) related to in-vivo counting include: 

• 	 Assumed mean body burdens of Cs-137 from fallout in the United States are provided in 

Table 5-16. ORNL used a value of 20 nCi Cs-137 for consumers of venison (pg. 26). 

• 	 Table 5-12 of the ORNL Internal Dose TBD provides the 1965 criteria for whole-body 

counting. This table does not exclude Cs-137 (pg. 21). ORNL uses a value of 20 nCi 

Cs-137 in a whole-body count for consumers of venison as the decision level to follow up 

or conduct a dose assessment (pg. 25). 

• 	 The ORNL Internal Dose TBD states, “It is a property of the Poisson distribution that its 

standard deviation is the square root of its mean value. Therefore, in this analysis, the 

standard deviation of the total background counts was calculated as the square root of the 

total background counts” (pg. 37). 

• 	 “The default radionuclide to use with measurements involving trivalent alpha 

actinides would be Am-241” (Bollenbacher et al. 2006, pg. 18). 

• 	 Table 5-15 summarized the maximum activity measured by in-vivo monitoring from 

1961 to 1966. The list of radionuclides includes Na-24, Sc-48, Cr-51, Co-56, Co-57, Co58, Fe-59, Co-60, Co-64, Zn-65, Se-75, Sr-90, Zr/Nb-95, Ru/Rh-106, Sb-125, I-131, Cs137, Eu-155, Hg-203, Ra-226, Pa-233, and enriched uranium (pg. 24). 

Assumptions discussed in the ORNL Occupational Internal Dose TBD (Bollenbacher et al. 2006) 

related to use of air sampling data include the following: 

• 	 ORNL maintained early tolerance levels for airborne contamination based on “product” 

(i.e., Pu-239) concentrations in the air (Cox 1944) (pg. 9). 
 

 

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• 	 Air-monitoring records for 1944 to 1947 can be used for assigning best estimates of dose 

(pg. 28). 

• 	 The basis for estimation of internal dose by air sampling is presented in Internal Dose 

Overestimates for Facilities with Air Sampling Programs (ORAU 2005) with ORNLspecific modifications. Alpha emitters are limited to uranium and plutonium. The dose 

is entered into IREP as a constant. Air-monitoring data prior to June 1944 is assumed. 

Airborne rates would be less than in later periods (pg. 28). 

• 	 Exposure to the maximum permissible concentration for 40 hours a week, 50 weeks per 

year (pg. 67). 

• 	 The radionuclide that results in the largest intake is assumed to comprise 100% of the 

intake. The radionuclides may change from year to year (pg. 67). 

• 	 The ORNL Internal Dose TBD did not discuss the establishment of boundary conditions 

for chronic intakes. 

• 	 In the ORNL Internal Dose TBD, Table 5A-2 in Attachment 5A, page 39, provides 

annual averages for periods of time where bioassay data were available. This table 

includes MDAs calculated from recovered data for radionuclides. The table shows data 

for Pu starting in 1946, for 2 or 3 others by 1949, and for 8 radionuclides by 1952. There 

is no discussion of applying 1951 data to earlier periods. 

External Exposure 

The ORNL Occupational External Dose TBD ORAUT-TKBS-0012-6 (Burns and Mohrbacher 

2004) describes the default assumptions for occupational external dose at ORNL. The dose 

assignments were based on pocket dosimeter, film badge, and thermoluminescent dosimeter 

results. Attachment 6A gives the IREP energy groupings for photon exposure by facility 

(pg. 76). Attachment 6B contains a list of ORNL facilities where workers may have received 

neutron exposure, their dates of operation, and the appropriate neutron energies (pg. 24). 

Attachment 6C contains a summary of neutron-to-photon ratios derived from field measurements 

(pg. 24), and is provided for information purposes primarily. Attachment 6D provides the 

neutron-to-photon ratios derived from dosimetry records as a function of worker group (pg. 45). 

Assumptions related to beta/gamma and neutron exposures include the following (Burns and 

Mohrbacher 2004): 

• 	 The ORNL dosimetry system became DOELAP-approved and Hp(10) doses were 

recorded. No exposure to organ-dose conversion factors is necessary (pg. 16). 

• 	 Claimant-favorable assumptions should be applied for radiation types and energies when 

converting to organ doses (pg. 22). 
 

 

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• 	 The recommended photon energies are > 250 keV (75%) and 30–250 keV (25%), except 

at certain accelerator and transuranic facilities where the shallow dose could be treated as 

100% <30 keV photon or 100% >15 keV electron. The choice should be made based on 

job responsibilities (pg. 22). 

• 	 Unlike the LANL External Dose TBD, which states that beta and photon energy 

distributions are determined by area (Widner 2005, pp. 69–70), the ONRL External Dose 

TBD does not discuss this. 

• 	 100% AP geometry is assumed for all workers (pg. 23). 

• 	 In the ORNL External Dose TBD, it states that “...dose reconstructors should apply 

claimant-favorable assumptions for radiation types and energies when converting deep 

and shallow beta-gamma dosimeter dose for ORNL workers to organ dose for input to the 

Interactive RadioEpidemiological Program (IREP)” (pg. 22). 

• 	 The ORNL External Dose TBD cites the following reference regarding the use of a factor 

of 2.5. “The purpose of the term 2.5 (W–P) in the expression for DS above was to 

represent the beta portion of the total dose, with the factor of 2.5 being a correction for 

the attenuation by the OW and plastic filters relative to a dose depth of 7 mg/cm2 . The 

factor of 2.5 was based on the beta spectrum from natural uranium, but different factors 

could be used if there was evidence of exposure to a different beta energy (Thornton, 

Davis, and Gupton 1961).” 

• 	 The ORNL External Dose TBD photon missed dose values in terms of LODs for deep 

dose and dosimeter exchange frequencies on ORNL radiation workers are to be done in 

accordance with Table 6-24 (pg. 69). The use of MDL/2 is not discussed. LODs and 

exchange frequencies are provided in Table 6-24 (pg. 69) for certain periods. 

• 	 Review of claims indicates that B values are the difference between OW and shielded 

readings; therefore, the results should be interpreted as such (pg. 60). 

• 	 Apply an organ dose conversion factor to the deep dose results for the period of 1989 

(pg. 67). 

• 	 Before 1961, sum the beta and gamma values to obtain the shallow dose (pg. 67). 

• 	 Multiply the OW values from 1944 to February 1947 by a factor of 2.2. Multiply the 

OW results by 0.9 from February 1947 through February 1951. These factors represent 

adjustment factors to account for calibration differences (pg. 67). 

• 	 If any reported neutron dose data are used, the fast neutron value should be multiplied by 

factor of 4 for 1943–1950 and a factor of 2 for 1950–1968 (pg. 68). 
 

 

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• 	 Missed shallow dose should apply the LOD values for deep dose unless the field consists 

of primarily low-energy beta. In this case, if Centralized External Dosimetry System 

(CEDS) TLD were used, the LOD was 14 mrem/quarter (pg. 69). 

• 	 Use neutron-to-photon ratios unless the NTA film yields a high dose. These ratios should 

be applied to both recorded and missed photon dose where neutron dose is anticipated 

(pg. 70). 

• 	 For reactors or other facilities not addressed in Attachments 6C or 6D, evaluate the 

applicability of the distribution of neutron-to-photon dose ratios derived from the ORNL 

personnel monitoring records for similar facilities or job duties, the distribution for all 

facilities, or the distribution yielding the maximum ratios (pg. 71). 

• 	 Uncertainty factors for reported ORNL doses are provided based on characterization of 

dosimeters by site personnel and the DOELAP analysis. The standard error used for 

beta/gamma film badge doses is estimated to be ± 30%. The error for a neutron dose 

with a 100 mrem reading is estimated at ± 50%. For dosimeters used after the 

mid-1970s, the standard errors for a 100 mrem exposure were approximately ±15%. 

• 	 In the ORNL External Dose TBD, in footnotes of both Attachment 6C (pg. 79) and 

Attachment 6D (pg. 80), both dealing with neutron-to-photon ratios at ORNL, states that 

the geometric mean, minimum, and maximum should be doubled to account for ICRP 60 

radiation weighting factors. 

• 	 Dose reconstructors should refer to the neutron-to-photon dose ratios in Attachments 6C 

and 6D to assign missed and unmonitored dose for individuals who received neutron 

exposures in comparable facilities or job duties (pg. 71). 

• 	 In the ORNL Site Description TBD, it is pointed out that neutron exposure potentially 

occurred at the accelerators, the Building 7735 Dosimetry Applications Research 

Calibration Laboratory (DOSAR), the ORNL Research Reactor in Building 3042 (ORR), 

the Tower Shielding Facility (TSF) in the 7700 area, the neutron flight tube facility 

(3083), the Neutron Users Office and Lab (7962), the Neutron Science Support Center 

(7970), isotope separations facilities, transuranic facilities, and waste storage and disposal 

facilities (pg. 77). 

• 	 Table 6-23 provides group-weighted NCRP 38 (NCRP 1971) quality factors and 

correction factors used to convert NCRP values to the appropriate ICRP 60 (ICRP 1990) 

value. Conversion factors are 2.1 for thermal neutrons, 1.9 for 0.01–2.0 MeV neutrons, 

and 1.3 for 2.0–14.0 MeV neutrons (pp. 68–69). 

• 	 For 1975 to the present, neutron missed dose can be estimated on the basis on LODs and 

dosimeter exchange frequencies (Burns and Mohrbacher 2004, pg. 71). 

• 	 The ORNL External Dose TBD provides the following guidance regarding the 

assignment of dose by work location: if possible: “....actual personnel neutron and 
 

 

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photon exposures and should be used in favor of the values given in Attachment 6C as 

practicable. The values in Attachment 6C are derived from characterization of workplace 

neutron spectra performed in conservatively chosen locations with high neutron dose 

rates” (pg. 71). 

• 	 A co-worker model is available for beta/gamma exposures; however, a method has not 

been developed for neutrons (Merwin 2004). 

The default assumptions for entry of individual dose into IREP are consistent with those used in 

other TBDs. As with other site profiles, the missed photon dose is determined by dividing the 

Minimum Dose Limit (MDL) by 2 and multiplying by the number of zeros and unmonitored 

periods. The missed photon dose is entered into IREP as a lognormal distribution with a 

geometric standard deviation of 1.52. Dosimeter correction factors are based on site-specific 

information, which is appropriate. 

Inconsistencies within the ORNL Site Profile 

The TBD states that it does not address skin exposure, yet it provides shallow dose correction 

factors. Furthermore, during the SC&A conference call with NIOSH, the dose reconstruction 

representatives indicated that the shallow dose is assigned using OTIB-0017 (see Attachment 4). 

This appears to be providing a methodology to assess skin exposure. Clarification on shallow 

dose calculation is needed in the TBD. 

The Site Description TBD identifies radionuclides by facility and includes information on 

incidents and spills that occurred in the facilities. Some, but not all, of this information is 

considered in the Environmental Dose TBD and the Internal Dose TBD. In particular, there are 

radionuclides (e.g., medical isotopes) that have not been included in the dose calculations. These 

radionuclides should be evaluated to determine their significance to environmental and internal 

dose. 

Inconsistencies between Site Profiles 

Some variation occurs between several of the TBDs with respect to the assumed chest thickness. 

The TBD (Fleming 2006a, Table 3-3, pg. 18) assumes a PA and lateral chest thicknesses of 26 

and 34 cm, respectively, at ORNL; AP and lateral LS thicknesses of 26 and 34 cm, respectively; 

and distance from body to imaging surface of 5. The SRS TBD (Scalsky 2005) assumes 26 cm 

and 34 cm for PA chest x-rays and lateral chest x-rays, respectively. The Hanford and ORNL 

TBDs actually recommend adjustment factors for chest thickness. There is no mention of 

applying a factor of 2.5 to the PA x-rays to estimate the ESE for lateral chest x-rays. This factor 

has been included in many of the other TBDs and OTIB-0006 (Kathren and Shockley 2005, 

pg. 20). 

The Occupational Medical Dose TBDs for Y-12 (Murray 2003), SRS (Scalsky 2005), and 

Hanford (Scalsky 2003) base their default exposure geometry on the compensability or noncompensability of the claim. The MCW (Westbrook 2005) and RFP (Furman and Lopes 2004) 

Occupational Medical Dose TBDs base default exposure geometries on job titles. Both the 
 

 

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ORNL TBD (Fleming 2006a) and the INEEL (Rohrig 2004) Occupational Medical Dose TBDs 

choose to default to 100% AP exposure. Further evaluation of exposure geometry for photon 

and neutron exposure should be evaluated for ORNL workers to determine if 100% AP geometry 

is appropriate for all ORNL workers. The NIOSH/ORAU team should consider development of 

a consistent default assumption for exposure geometry in all site profiles. 

The ORNL Environmental Dose TBD assumes that average annual site-wide values are assigned 

to unmonitored workers. For LANL and SRS, when the worker location is unknown, the sitewide maximum ambient radiation dose is assigned (Cehn and McDowell-Boyer 2004; Scalsky 

2005). The application of multiple receptor points would be consistent with the approach in the 

Hanford, SRS, and INEEL TBDs. The use of measured air concentration data is consistent with 

the Y-12 Occupational Environmental Dose TBD (Ijaz and Adler 2004). The ORNL TBD 

(Burns 2004b) appears to rely on emission and measurement data; however, it does not indicate 

the model used for calculations. The TBD generally discusses particle size; however, the actual 

particle size assumptions for assignment of internal dose have not been provided. 

The ORNL TBD assumes exposures from I-131, H-3 (starting 1967), Kr-85, Xe-133, and MFPs 

(optional). Other site profiles have considered radionuclides hat were present at ORNL. The 

Hanford TBD included plutonium, noble gases, and MFPs (Savignac 2003). The SRS TBD 

included tritium, noble gases, plutonium, and uranium (Scalsky 2005). The LANL TBD 

included tritium, noble gases, plutonium, americium, uranium, and thorium (Cehn and 

McDowell-Boyer 2004). Several of the ORNL workers were stationed at the Y-12 Plant 

location, indicating these radionuclides may not be applicable. Y-12 Plant handled primarily 

uranium, with some processing of thorium and U-233 (SC&A 2005). These radionuclides have 

not been considered in the assignment of dose to workers at the Y-12 facility. 

With the LANL Environmental Dose TBD, a screening method was applied to the source term to 

determine relative dose consequences for different radionuclides. With the similarities in the 

breadth of radionuclides handled at ORNL and LANL, this screening method or equivalent is 

considered prudent at ORNL. Overall, SC&A believes that further investigation into 

environmental source terms is needed. 

The ORNL Internal Dose TBD lacks guidance on the activity fractions for plutonium and 

thorium. For example, SRS assumes a 10-year old, 12% plutonium mixture (Scalsky 2005). 

Activity fractions for plutonium provide critical information for the assessment of dose from 

americium as an impurity. Americium doses have been considered in other site profiles. The 

ORNL TBD is silent about exposures from recycled uranium (RU), although it is considered in 

the Y-12 TBD (Rich and Chew 2006). The Y-12 site supplied materials to ORNL and had 

individuals stationed at the Y-12 Plant site. There is no mention of whether special tritium 

compounds were handled at ORNL, and how intakes from this material are to be treated. Further 

investigation should be conducted related to ORNL employee exposure to RU and tritides. 

There is mention of the presence of the release radon from stacks and storage areas; however, no 

discussion was included for potential internal dose to radon. It would seem reasonable that radon 

would be present in the workplace if it were being released to the environment. Radium-226 and 

ores were also handled at ORNL. Radon was specifically mentioned in the MCW and Fernald 
 

 

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TBDs, where high concentrations of Ra-226 were handled. The ORNL Site Profile fails to 

consider radon exposure. The impact of radon on dose reconstruction should be evaluated, 

particularly for the early years. 

During the Y-12 Special Exposure Cohort petition review, considerable attention was given to 

radionuclides generated by the ORNL Isotope Production Group and potential exposures to Y-12 

workers from these activities. ORNL employees were stationed at the Y-12 Plant and actively 

ran these operations. The radionuclides of concern for Y-12 Plant are H-3, Sr-90, Tc-99, 

Th-228, Th-232, U-232, U-233, U-236, Pu-238, Pu-239, Pu-241, Np-237 Am-241, Co-60, 

Zr/Nb-95, and Po-208. This is a somewhat different mixture of radionuclides than those 

considered for missed dose at the ORNL main site. The Internal Dose TBD (Bollenbacher et al. 

2006) does not include a discussion on accelerator-produced radionuclides, or medical isotope 

production and how these are handled in dose reconstruction in terms of missed dose. 

There has been no consideration of ingestion dose for ORNL, particularly when air sampling 

data is used to determine internal dose. The Bethlehem Steel and MCW TBDs included 

ingestion as a potential route of internal dose. This may be particularly important at LANL in 

the early years. 

The ORNL Internal Dose TBD does not refer the dose reconstructor to the specific guidance 

provided in OCAS-TIB-002 (OCAS 2003) when the dose records indicate an exposure to 

organically bound tritium (OBT) and metal tritides (MTs). This referral was made in the LANL 

Internal Dose TBD (Argall 2004). The Mound Internal Dose TBD (Millard 2004), in 

Section 5.3.1.1, Metal Tritides, states that a lung clearance class of S should be assumed for all 

metal tritides other than lithium. The SRS Internal Dose TBD and associated TIBs fail to treat 

the topic of dose reconstruction from exposure to organically bound tritium and metal tritides. 

Consideration should be given to identifying those facilities with special tritium compounds and 

developing a common methodology to assign dose from the compounds. 

In general, the Integrated RadioEpidemiology Program (IREP) input criteria for Radiation Rate, 

Radiation Type, and Dose Distribution Type are the same for ORNL, as for other site profiles. 

Maximizing internal dose is assigned with the use of ORAUT-OTIB-0018, Internal Dose 

Overestimates for Facilities with Air Sampling Programs; or ORAUT-OTIB-0002, Technical 

Information Bulletin: Maximum Internal Dose Estimates for Certain DOE Complex Claims. 

The application of these OTIBs for the assignment of dose is consistent with other site profiles. 

External dose assumptions were consistent with those used in other site profiles. The default 

energies for beta, photon, and neutron exposure were >15 keV, 30–250 keV, and 0.1–2.0 MeV, 

respectively. The missed external dose is calculated using the MDL/2 calculation times the 

number of monitoring periods, and is entered as a lognormal distribution with a geometric 

standards deviation of 1.52. This is consistent with other site profiles. 

The ORNL Occupational External Dose TBD provided Attachment 6A on page 76 to allow the 

dose reconstructor’s selection of IREP energy groups by process location for beta-gamma 

exposures. Likewise, Attachment 6B on page 78 provides similar guidance by process location 

for neutron-to-photon ratios at ONRL facilities for 100 % occupancy. The LANL Occupational 

External Dose TBD (Widner 2005) applies an area-specific neutron energy distribution for years 
 

 

 

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after 1978. For years prior to 1980, the annual photon dose is multiplied by a neutron-to-photon 

ratio to obtain the neutron dose. The TBD (Widner 2005) has derived three neutron-to-photon 

ratios for the plutonium facilities, criticality experiments (> 50 m distant), and other operations. 

The SRS TBD distinguishes neutron energies and neutron-to-photon ratios for reactors, fuel 

fabrication, plutonium production, and radionuclide production and calibration (Scalsky 2005). 

The INEEL Occupational External Dose TBD (Rohrig 2004) considers the reactors, the 

processing plant, waste-handling operations, calibration sources, and uranium handling. Neutron 

energy specta and neutron-to-photon ratios for Pu-238 and Pu-239 operations are segregated at 

SRS (Scalsky 2005). The categories used in the LANL TBD (Widner 2005) lack the detailed 

analyses seen in other TBDs. Further evaluation of neutron-to-photon ratios should include more 

specific categories, including neutron sources (RaBe, Cf, etc.), accelerators, early subcriticality 

experiments, initiator development, and neutron spectra from alternate fissile materials. 

There also is no mention of whether the background uranium is subtracted from the urinary 

bioassay results in the ORNL Occupational Internal Dose TBD (Bollenbacher et al. 2006, 

pg. 19), but the LANL Occupational Internal Dose TBD (Argall 2004, pg. 29) states that the 

background uranium is subtracted from the bioassay results. SC&A is concerned with what 

appears to be an inconsistency in how the excretion of naturally occurring uranium is handled in 

the dose reconstruction between site profiles. 

There is discussion in the Occupational Internal Dose TBD (Bollenbacher et al. 2006) of the 

levels of naturally occurring environmental uranium and the ratios of isotopes. However, it is 

not clear how NIOSH is going to handle this with regard to levels seen in urinary and fecal 

excretion and the bioassay monitoring data. Section 5.2.3.3, page 19, states that, “Plots of the 

observed uranium excretion distributions for U-234, U-238, and total uranium are provided 

below.” SC&A, however, fails to find these data in the document. There also is no mention of 

whether the background uranium is subtracted from the urinary bioassay results in the ORNL 

Occupational Internal Dose TBD (Bollenbacher et al. 2006, pg. 19), but the LANL Occupational 

Internal Dose TBD (Argall 2004, pg. 29) states that the background uranium is subtracted from 

the bioassay results. SC&A is concerned with what appears to be an inconsistency in how the 

excretion of naturally occurring uranium is handled in the dose reconstruction between site 

profiles. 

SC&A is likewise concerned with the conflicting assumptions for solubility classification of 

uranium in between the ORNL Site Profile (Bollenbacher et al. 2006) and the LANL Site Profile 

(Argall 2004). The ORNL TBD states that the default to be used in dose reconstruction should 

be Type S (Table 5-2). The LANL Internal Dose TBD (Argall 2004) recommended Type M. 

The ORNL Internal Dose TBD does not address the Integrated Modules for Bioassay Analysis 

(IMBA) NIOSH Phase I database USDOE /Version 1.0.42, Tables 5-11 and 5-12, for isotopic 

composition for uranium. SC&A is concerned over some of these inconsistencies shown by 

NIOSH between site profiles. 

The External Dose TBDs for Y-12 (Kerr 2006), SRS (Scalsky 2005), and Hanford (Fix 2004) 

base their default exposure geometry on the compensability or noncompensability of the claim. 

The MCW (Westbrook 2005) and RFP (Furman and Lopez 2004) TBDs based the default 

exposure geometries on job titles. The LANL (Widner 2005), ORNL, and INEEL (Rohrig 2004) 
 

 

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TBDs default to 100% AP geometry. Further evaluation of the exposure geometry for photon 

and neutron exposure should be evaluated for ORNL workers to determine if this is appropriate 

for all ORNL workers. 

The ORNL External Dose TBD does not provide clear guidance on when to assign missed 

neutron dose, except to recommend that it be assigned when neutron exposure is expected. In 

the LANL, SRS, Hanford, and additional site profiles, the basis for assigning missed neutron 

dose is the work location and/or the existence of neutron dosimetry during another period of 

time. In the case of LANL, the neutron exposures from co-workers were also considered an 

indicator of potential neutron exposure (Widner 2005). 

The current Occupational External Dose TBD (Burns and Mohrbacher 2004) states it does not 

address reconstruction of skin dose. Shallow dose at ORNL is determined by the procedure 

outlined in ORAUT-0TIB-0017 (see Attachment 3). 

The LANL External Dose TBD (Widner 2005) has developed facility-specific neutron energy 

distributions and neutron-to-photon ratios. Furthermore, neutron-to-photon ratios have been 

developed for worker groups that received neutron exposure. The SRS TBD distinguishes 

neutron energies and neutron-to-photon ratios for reactors, fuel fabrication, plutonium 

production, and radionuclide production and calibration (Scalsky 2005). The INEEL 

Occupational External Dose TBD (Rohrig 2004) considers the reactors, the processing plant, 

waste-handling operations, calibration sources, and uranium handling. Neutron energy spectra 

and neutron-to-photon ratios for Pu-238 and Pu-239 operations are segregated at SRS (Scalsky 

2005). 
 

 

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References 

Argall, Rowena S., 2004, Technical Basis Document for the Los Alamos National Laboratory – 

Occupational Internal Dose, ORAUT-TKBS-0010-5, Rev. No. 00, December 21, 2004. 

Cox, A.C., 1944, Progress Report – First Nine Months Operation of the Pilot Plant, CN-1879, 

October 9. 

Fix, Jack, 2004, Technical Basis Document for the Hanford Site –Occupational External 

Dosimetry, ORAUT-TKBS-0006-6, Rev. No. 01, January 9, 2004. 

ORAU 2005, Internal Dose Overestimates for Facilities with Air Sampling Programs, ORAUT0TIB-0018, Rev. 01, Oak Ridge Associated Universities, Oak Ridge, Tennessee, August 9, 2005. 

Rich, Bryce L. and Chew, M.H., 2006, Y-12 National Security Complex –Occupational Internal 

Dose, ORAUT-TKBS-0014-5, Rev. 02, August 3, 2006. 

Thornton, W.T., Davis, D.M., and Gupton, E.D., 1961, The ORNL Badge Dosimeter and its 

Personnel Monitoring Applications, ORNL-3126, Oak Ridge National Laboratory, Oak Ridge, 

Tennessee.